[Federal Register Volume 60, Number 128 (Wednesday, July 5, 1995)]
[Notices]
[Pages 35058-35091]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-16249]



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NUCLEAR REGULATORY COMMISSION

Applications and Amendments to Facility Operating Licenses 
Involving No Significant Hazards Considerations; Biweekly Notice

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from June 10, 1995, through June 22, 1995. The 
last biweekly notice was published on June 21, 1995 (60 FR 32359).

Notice of Consideration of Issuance of Amendments to Facility Operating 
Licenses, Proposed No Significant Hazards Consideration Determination, 
and Opportunity for a Hearing

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. 

[[Page 35059]]
Under the Commission's regulations in 10 CFR 50.92, this means that 
operation of the facility in accordance with the proposed amendment 
would not (1) involve a significant increase in the probability or 
consequences of an accident previously evaluated; or (2) create the 
possibility of a new or different kind of accident from any accident 
previously evaluated; or (3) involve a significant reduction in a 
margin of safety. The basis for this proposed determination for each 
amendment request is shown below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Rules Review and 
Directives Branch, Division of Freedom of Information and Publications 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555, and should cite the publication date and page 
number of this Federal Register notice. Written comments may also be 
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the NRC Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC. 
The filing of requests for a hearing and petitions for leave to 
intervene is discussed below.
    By August 4, 1995, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington DC, by the above 
date. Where petitions are filed during the last 10 days of the notice 
period, it is requested that the petitioner promptly so inform the 
Commission by a toll-free telephone call to Western Union at 1-(800) 
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator 
should be given Datagram Identification Number N1023 and the 

[[Page 35060]]
following message addressed to (Project Director): petitioner's name 
and telephone number, date petition was mailed, plant name, and 
publication date and page number of this Federal Register notice. A 
copy of the petition should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and 
to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit Nos. 
1, 2, and 3, Maricopa County, Arizona

    Date of amendment requests: May 2, 1995.
    Description of amendment requests: The proposed amendment would 
remove from the technical specifications (TS) plant elevations for the 
minimum water volume required in the spent fuel pool (SFP) and relocate 
them to site procedures. This proposed TS amendment also includes two 
changes to correct administrative errors in the TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis about the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change eliminates the plant elevations from TS 
Figure 3.1-1, ``Minimum Borated Water Volumes'' for the SFP. The 
change is administrative in nature and does not involve any 
modifications to plant equipment or affected plant operation. The 
required volume of water in the SFP is identified on the figure and 
will remain unchanged by this amendment. This request relocates the 
plant elevations to site procedures where they will be controlled in 
accordance with the provisions of 10 CFR 50.59.
    The removal of the reference to Table 3.8-2 in the Unit 3 TS 
3.8.4.1 and adding the word ``containment'' to the Unit [2] TS 
4.6.3.1 are administrative change[s] and do not involve any 
modifications to plant equipment or affect plant operation. These 
administrative changes do not affect the scope or intent of any test 
within the TS.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change eliminates the plant elevations from TS 
Figure 3.1-1, ``Minimum Borated Water Volumes'' for the SFP. The 
change is administrative in nature and does not involve any 
modifications to plant equipment or affect plant operation. The 
removal of plant elevations from the figure does not cause any 
change in the method by which any safety-related system performs its 
function. The required volume of water in the SFP is identified on 
the figure and will remain unchanged by this amendment.
    The removal of the reference to Table 3.8-2 in the Unit 3 TS 
3.8.4.1 and adding the word ``containment'' to the Unit 2 TS 4.6.3.1 
are administrative changes and do not involve any modifications to 
plant equipment or affect plant operation. These administrative 
changes do not affect the scope or intent of any test within the TS.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change eliminates the plant elevations from TS 
Figure 3.1-1, ``Minimum Borated Water Volumes'' for the SFP. The 
change is administrative in nature and does not involve any 
modifications to plant equipment or affect plant operation. The 
required volume of water in the SFP is identified on the figure and 
will remain unchanged by this amendment.
    The removal of the reference to Table 3.8-2 in the Unit 3 TS 
3.8.4.1 and adding the word ``containment'' to the Unit 2 TS 4.6.3.1 
are administrative changes and do not involve any modifications to 
plant equipment or affect plant operation. These administrative 
changes do not affect the scope or intent of any test within the TS.
    Therefore, based upon the above, the proposed change does not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
that review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004.
    Attorney for licensees: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999.
    NRC Project Director: William H. Bateman.

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318, 
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, 
Maryland

    Date of amendments request: June 2, 1995.
    Description of amendments request: The proposed amendments would 
revise the pressurizer safety valve setpoint tolerance ``as-found'' 
acceptance criterion to +2%/-1% for the valve with the lower setpoint 
(RC-200) and plus or minus 2% for the valve with the upper setpoint 
(RC-201). The ``as-left'' setpoint tolerance will remain plus or minus 
1% for both valves.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The pressurizer safety valves are used to prevent exceeding the 
Reactor Coolant System (RCS) pressure safety limit. The proposed 
change to increase the pressurizer safety valve setpoint tolerance 
for the ``as-found'' acceptance criteria from [plus or minus]1% to 
+2%/-1% for the valve with the lower pressure setpoint, and [plus or 
minus] 2% for the valve with the upper pressure setpoint, does not 
affect any initiating event. The proposed change does not affect the 
consequences of the previously evaluated design basis accidents as 
the new safety valve setpoint tolerances are bounded by the 
assumptions in the safety analysis. Therefore, the proposed change 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Would not create the possibility of a new or different type 
of accident from any accident previously evaluated.
    The proposed change to increase the ``as-found'' setpoint 
tolerances does not involve any changes in equipment or the function 
of these safety valves. The proposed change does not represent a 
change in the configuration or operation of the plant. The test 
method for the pressurizer safety valves will remain the same. The 
increase in the setpoint tolerances does not create any new accident 
initiator. Therefore, the proposed change does not create the 
possibility of a new or different type of accident from any accident 
previously evaluated. 

[[Page 35061]]

    3. Would not involve a significant reduction in a margin of 
safety.
    The pressure safety limit for the RCS protects the structural 
integrity of the system from failure due to overpressurization. The 
pressurizer safety valves are used to prevent the RCS pressure from 
exceeding the safety limit. The proposed change to the pressurizer 
safety valve setpoint tolerances will continue to prevent the RCS 
pressure from exceeding the design safety limit during any design 
basis event. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.
    Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Ledyard B. Marsh.

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318, 
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, 
Maryland

    Date of amendments request: June 6, 1995.
    Description of amendments request: The proposed amendments would 
revise the Calvert Cliffs Nuclear Plant Units 1 and 2 Technical 
Specifications, extending certain 18-month frequency surveillances to a 
refueling interval (nominally 24 months, not to exceed 30 months). 
Systems and equipment affected are the Reactor Protective System (RPS), 
Engineered Safety Features Actuation System (ESFAS), Power-Operated 
Relief Valve (PORV) actuation instruments, Low Temperature Overpressure 
Protection (LTOP)-related instruments, Remote Shutdown Panel 
instruments, Post-Accident Monitoring (PAM) instruments, Containment 
Sump Level instruments, and Radiation Monitoring instruments.
    This amendment request would extend the nominal surveillance 
interval requirement from 18 months to a refueling interval (nominally 
24 months, not to exceed 30 months) for instrument channel 
calibrations, RPS and ESFAS total bypass function operability 
verification, RPS and ESFAS time response tests, ESFAS Manual Trip 
Button channel functional tests, and ESFAS Automatic Actuation Logic 
Channel Functional Tests. Calvert Cliffs has been operating on a 24-
month fuel cycle since July 1987 (Unit 2) and July 1988 (Unit 1), 
performing some Technical Specification surveillances, such as the ones 
described here, during mid-cycle outages. The request is the last of a 
series of proposed license amendments that would eliminate the need for 
planned mid-cycle outages to perform required surveillances.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The proposed change would extend surveillance intervals for 
Reactor Protective System (RPS), Engineered Safety Features 
Actuation System (ESFAS), Power-Operated Relief Valve (PORV), Low 
Temperature Overpressure Protection (LTOP), Remote Shutdown, Post-
Accident Monitoring (PAM), Radiation Monitoring, and Containment 
Sump Level Instruments.
    The purpose of the RPS is to effect a rapid reactor shutdown if 
any one or a combination of conditions deviates from a pre-selected 
operating range. The system functions to protect the core and the 
Reactor Coolant System pressure boundary. The purpose of the ESFAS 
is to actuate equipment which protects the public and plant 
personnel from the accidental release of radioactive fission 
products if an accident occurs, including a loss-of-coolant 
incident, main steam line break, or loss of feedwater incident. The 
safety features function to localize, control mitigate, and 
terminate such incidents in order to minimize radiation exposure to 
the general public. The Post-Accident Monitoring instruments provide 
the Control Room operators with primary information necessary to 
take manual actions, as necessary, in response to design basis 
events, and to verify proper system response to plant conditions and 
operator actions. The purpose of the Remote Shutdown System is to 
provide plant parameter indications to operators on a Remote 
Shutdown Panel to be used while placing and maintaining the plant in 
a safe shutdown condition in the event the Control Room is 
uninhabitable. The indications are used to verify proper system 
response to plant conditions and operator actions. The LTOP System 
protects against Reactor Coolant System overpressurization at low 
temperatures by a combination of administrative controls and 
hardware. The hardware includes two Power-Operated Relief Valves 
with variable pressurizer pressure setpoints when operating in the 
LTOP operating parameter region. The Containment Sump High Level 
Alarm System provides an alarm in the Control Room for a containment 
sump to provide one of the available indications of excessive RCS 
leakage during normal plant operation. The Containment Area High 
Range Radiation Monitoring System provides an indication of high 
radiation levels in containment. The Containment Purge System 
actuates equipment to prevent the release of radioactive material to 
the environment in the event of a reactor coolant leak, a shielding 
failure, or a fuel pin failure when the reactor vessel head is 
removed.
    The instruments in each of the systems described above are 
designed to be used in response to an accident. Failure of any of 
these systems is not an initiator for any previously evaluated 
accident. Therefore, the proposed change would not involve an 
increase in the probability of an accident previously evaluated.
    Many of the instruments addressed in this license amendment 
request will have or have recently had a new brand of sensor 
installed. The effect of the increased surveillance interval with 
the new sensors was analyzed. The new sensors do not effect the 
physical design description of the plant, any design or functional 
requirements, or surveillances. The proposed Technical Specification 
change extending the surveillance interval from 18 months to a 
refueling interval (nominally 24 months, not to exceed 30 months) 
does not physically change the plant, change any design or 
functional requirements, or effect the surveillances themselves. 
Analysis has shown that no trip setpoints need to be changed, and 
operator indications will continue to be accurate for control of 
plant parameters to effect a safe shutdown. The equipment will 
continue to perform as designed to mitigate the consequences of 
accidents. Therefore, the proposed change would not involve a 
significant increase in the consequences of an accident. [* * *]
    Therefore, the proposed change would not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Would not create the possibility of a new or different type 
of accident from any accident previously evaluated.
    The proposed change to increase the interval RPS, ESFAS, PORV, 
LTOP, Remote Shutdown, PAM, Radiation Monitoring, and Containment 
Sump Level instrument surveillances from 18 months to a refueling 
interval (nominally 24 months, not to exceed 30 months) does not 
involve a significant change in the design or operation of the 
plant. No hardware is being added to the plant as part of the 
proposed change. Some detector upgrades in specific plant systems to 
enhance the performance of those systems have been or will be 
performed. However, those upgrades were evaluated and deemed 
acceptable under 10 CFR 50.59 and are not part of this request. The 
Reactor Protective System, Engineered Safety Features Actuation 
System, Power-Operated Relief Valve, Low Temperature Overpressure 
Protection, Containment Sump Level, one Radiation Monitoring 
actuation setpoints will not be changed. Analysis has shown that the 
remote shutdown and PAM indications will continue to be accurate. 
The proposed change will not introduce any new accident initiators. 
Therefore, this change does not create the possibility of a new or 
different type of accident from any previously evaluated.
    3. Does operation of the facility in accordance with the 
proposed amendment 

[[Page 35062]]
involve a significant reduction in a margin of safety?
    The impact of the surveillance interval extension request was 
evaluated for each Technical Specification-related safety function 
for each of the RPS, ESFAS, PORV, LTOP, Remote Shutdown, PAM, 
Radiation Monitoring, and Containment Sump Level instruments 
addressed by this submittal. In all cases, parameters specified in 
the related accident analysis were determined to be unaffected by 
the surveillance interval extension, and no accident analyses limits 
required changes. The Reactor Protective System, Engineered Safety 
Features Actuation System, Power-Operated Relief Valve, Low 
Temperature Overpressure Protection, Containment Sump Level, and 
Radiation Monitoring actuation setpoints will not be changed. 
Analysis has shown that the remote shutdown and PAM indications will 
continue to be accurate. The methods for detection of degraded 
instrument operation have not been changed, and remote shutdown and 
PAM operator indications will continue to provide adequate accuracy. 
The methods for detection of degraded instrument operation have not 
been changed, and remote shutdown and PAM operator indications will 
continue to provide adequate accuracy.
    The proposed change does not affect the operation of the systems 
involved. The surveillance interval extension will not affect the 
design of the systems, and methods for detection of degraded 
instrument operation will continue to identify operation problems 
between calibrations. Therefore, the proposed change does not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.
    Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Ledyard B. Marsh.

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318, 
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, 
Maryland

    Date of amendments request: June 9, 1995.
    Description of amendments request: The proposed amendments revise 
the Calvert Cliffs Nuclear Power Plant Radiological Effluent Technical 
Specifications (RETS) consistent with Generic Letter (GL), 
``Implementation of Programmatic Controls For Radiological Effluent 
Technical Specifications in the Administrative Controls Section of the 
Technical Specifications and the Relocation of Procedural Details of 
RETS to the Offsite Dose Calculation Manual or the Process Control 
Program (Generic Letter 89-01),'' dated January 31, 1989, and the 
Improved Standard Technical Specifications for Combustion Engineering 
Plants published in NUREG-1432, as modified by Mr. W. T. Russell's 
letter of October 25, 1993, ``Content of Standard Technical 
Specifications,'' to the Improved Technical Specification Owners Group 
Chairpersons. Changes for relocating the procedural details of the 
current RETS to the Offsite Dose Control Manual (ODCM) has been 
prepared in accordance with the proposed changes to the Administrative 
Controls section of the Technical Specifications.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The proposed change has been evaluated against the standards in 
10 CFR 50.92 and has been determined to not involve a significant 
hazards consideration, in that operation of the facility in 
accordance with the proposed amendments:
    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The proposed changes will provide human factor improvements for 
the Technical Specifications by relocating existing procedural 
details of the current Radiological Effluent Technical 
Specifications to the Offsite Dose Control Manual (ODCM). Procedural 
details for solid radioactive wastes will be relocated to the 
Process Control Program. The proposed amendment (1) incorporates 
programmatic controls in the Administrative Controls section of the 
Technical Specifications that satisfy the requirements of 10 CFR 
20.1302, 40 CFR Part 190, 10 CFR 50.36a, 10 CFR Part 50, Appendix I, 
and our current Technical Specifications; (2) relocates the existing 
procedural details in current specifications involving radioactive 
effluent monitoring instrumentation, the control of liquid and 
gaseous effluents, equipment requirements for liquid and gaseous 
effluents, radiological environmental monitoring, and radiological 
reporting details from the Technical Specifications to the ODCM; (3) 
simplifies the associated reporting requirements; (4) simplifies the 
administrative controls for changes to the ODCM; and (5) updates the 
definitions of the ODCM consistent with these changes.
    Relocating existing requirements and eliminating requirements 
which duplicate regulatory requirements provide Technical 
Specifications which are easier to use. Because existing 
requirements are relocated to established programs where changes to 
those programs are controlled by regulatory requirements, there is 
no reduction in commitment and adequate control is still maintained. 
Likewise, the elimination of requirements which duplicate regulatory 
requirements enhances the usability of the Technical Specifications 
without reducing commitments. The additional improvements being 
proposed neither add nor delete requirements, but merely clarify and 
improve the readability and understanding of the Technical 
Specifications. Since the requirements remain the same, these 
changes only affect the method of presentation, and as such, would 
not affect possible initiating events for accidents previously 
evaluated or any system functional requirement.
    Furthermore, no safety-related equipment, safety function, or 
plant operation will be altered as a result of this proposed change. 
The changes are unrelated to the initiation and mitigation of 
accidents and equipment malfunctions addressed in the Updated Final 
Safety Analysis Report.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Would not create the possibility of a new or different type 
of accident from any accident previously evaluated.
    Transferring the procedural details of radiological effluent 
monitoring and reporting from the Technical Specifications to the 
ODCM has no impact on plant operation or safety. No safety-related 
equipment, safety function, or plant operation will be altered as a 
result of this proposed change. No changes to plant components or 
structures are introduced which could create new accidents or 
malfunctions not previously evaluated.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Would not involve a significant reduction in a margin of 
safety.
    The margin of safety associated with the affected Technical 
Specifications is to provide assurance that the releases of 
radioactive materials during actual or potential releases of liquid 
or gaseous effluents do not exceed the limits of 10 CFR Part 20. 
This license amendment request relocates the methodology and 
parameters used to ensure that the 10 CFR Part 20 limits are 
maintained, but does not change any of these requirements. Thus, no 
methodology and parameters for controlling radioactive effluent 
releases will be changed.
    The procedural details of the current Radiological Effluent 
Technical Specifications will be transferred to the ODCM and 
replaced with programmatic controls consistent with regulatory 
requirements, including controls on revisions to the ODCM. Thus, no 
requirements or controls will be reduced.
    The proposed revisions to the reporting requirements for 
Radiological Effluent Release Report and the revision from the old 
10 CFR 20.106 requirements to the new 10 CFR 20.1302 have no impact 
on plant systems, plant operations or accident precursors. The 
changes to the effluent 

[[Page 35063]]
reporting requirements and the updated reference to 10 CFR 20.1302 do 
not change either the means of controlling radioactive releases or 
the effluent release limits. Therefore, there will be no change in 
the types and amounts of effluents that will be released, nor will 
there be an increase in individual or cumulative radiation exposures 
to any member of the public.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.
    Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Ledyard B. Marsh.

Carolina Power & Light Company, Docket No. 50-261, H.B. Robinson Steam 
Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of amendment request: June 3, 1995.
    Description of amendment request: The requested Technical 
Specification (TS) change clarifies the definition of operability of 
the charging pumps by adding a footnote to TS Section 3.2.2.a that 
states that the connectibility of the emergency power sources is not 
required for charging pump operability.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    This change request does not involve a significant hazards 
consideration for the following reasons.
    1. The requested change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The requested change clarifies that the emergency power 
sources are not required for the operability of the charging pumps. 
Operation of the charging pumps is not considered in the assumptions 
for initiation of any analyzed accident and is not credited for 
accident mitigation in any analyzed accidents in the safety analysis 
report. Therefore, the availability of emergency power sources to 
the charging pumps does not affect the probability of occurrence or 
consequences of an analyzed accident in the safety analysis report.
    2. The requested change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. The requested change clarifies that the emergency power 
sources are not required for the operability of the charging pumps. 
The design requirements of the charging pumps to provide reactor 
coolant inventory and boron inventory control are not changed. The 
operability of the emergency power source to the charging pumps is 
not a precursor to any accident scenario. Failure of the charging 
pumps is bounded by the plant design which strips the charging pumps 
from the emergency buses under certain conditions. Since the change 
does not involve changes in the operation of the plant, or physical 
or equipment changes or involve controls for accident mitigation 
equipment, the requested change will not create the possibility of 
new or different kind of accident from any accident previously 
evaluated.
    3. The requested change clarifies that the emergency power 
sources are not required for the operability of the charging pumps. 
Since the charging pumps are stripped from the emergency buses in 
the event of a loss of power and safety injection, emergency power 
sources to the charging pumps are not guaranteed to mitigate the 
consequences of an analyzed accident. As a result, no credit is 
taken for the charging function in analyzed accidents and the margin 
of safety as described in the safety analysis report is unchanged. 
Therefore, the requested change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550.
    Attorney for licensee: R. E. Jones, General Counsel, Carolina Power 
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602.
    NRC Project Director: David B. Matthews.

Commonwealth Edison Company, Docket Nos. 50-454 and 50-455, Byron 
Station, Unit Nos. 1 and 2, Ogle County, Illinois

Docket Nos. 50-456 and 50-457, Braidwood Station, Unit Nos. 1 and 2, 
Will County, Illinois

    Date of amendment request: February 21, 1995.
    Description of amendment request: The proposed amendments would 
revise Byron and Braidwood technical specifications associated with the 
reactor coolant system (RCS) resistance temperature detectors (RTDs) 
used to obtain hot and cold leg temperatures. The amendments are 
required because of proposed modification that will remove the existing 
RTDs and their associated piping and valves and replace them with dual 
element fast response RTDs mounted in the thermowells welded directly 
in the RCS loop piping.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed modification replaces the existing bypass piping 
system with thermowell-mounted RTDs. Because the hot leg RTDs are 
mounted directly in the scoops, temperature measurement inaccuracies 
caused by imbalances in the flow scoop sample flow are eliminated. 
The method of measuring coolant temperature with thermowell-mounted 
fast response RTDs has been analyzed to be at least as effective as 
the RTD bypass system. With the thermowells welded into the existing 
RCS hot and cold leg nozzles and the elimination of the bypass 
piping, the number of pressure boundary welds has been significantly 
reduced, resulting in a reduced probability of a small break LOCA 
[Loss of Coolant Accident].
    The RTD response time is incorporated in the safety analyses. In 
particular, RTD response time is modeled in the OT[DELTA]T [Over 
Temperature Delta Temperature] and OP[DELTA]T [Over Pressure Delta 
Temperature] trip functions. The overall response time modeled in 
the safety analyses for the existing RTD bypass piping system is 8 
seconds. The overall response time is the elapsed time from the time 
the temperature change in the RCS exceeds the trip setpoint until 
the rods are free to fall. More specifically, 6 seconds is modeled 
as a first order lag term and 2 seconds as pure delay on the reactor 
trip signal. The 6 second lag term includes such factors as: RTD 
bypass piping fluid transport delay, RTD bypass piping thermal lag, 
RTD response time, and RTD electronic filtering. The 2 second delay 
on reactor trip addresses such factors as electronics delay, trip 
breakers and gripper release.
    Signal conditioning (filtering) of the individual loop [DELTA]T 
and Tavg signals is represented by [time constants utilized in 
the lag compensator for DELTA T] and [time constant utilized in the 
measured Tavg lag compensator], respectively, in the OT[DELTA]T 
and OP[DELTA]T equations in Technical Specification Table 2.2-1. 
With the current bypass manifold system, the filter is not required 
since the existing RTDs do not respond rapidly to local temperature 
variances within the reactor coolant loop. The bypass piping and 
manifold provide adequate mixing of the coolant, eliminating any 
local temperature variances. Therefore, the values of [time 
constants utilized in the lag compensator for DELTA T] and [time 

[[Page 35064]]
constant utilized in the measured Tavg lag compensator] are 
currently specified as 0 seconds, effectively turning off the 
electronic filter. The new fast response RTDs may respond to 
temperature spikes which are not representative of actual RCS bulk 
fluid temperature. Signal conditioning may be required to eliminate 
these temperature spikes. Although, the current Technical 
Specifications do not provide for any signal conditioning, the 8 
second total response time used in safety analyses has sufficient 
margin to account for a typical 2 second time constant for signal 
conditioning. Industry experience has shown that a 2 second filter 
is adequate in eliminating the spikes.
    The proposed fast response RTD/thermowell system also has an 
overall response time of 8 seconds. However, the time distribution 
for the parameters differ between the existing and proposed designs. 
The existing design includes a transport time for RCS fluid to reach 
the RTD, located in the manifold. The RTDs are directly immersed 
into the coolant, providing a fast response. The new design no 
longer has the transport delay. However, because the RTDs are 
mounted in thermowells, the response time of the RTD/thermowell 
combination will be increased over the existing system.
    The effects of a redistribution of the time responses between 
the total lag term (pipe transport delay, RTD response and 
electronic filter delay) and electronics delay term have been 
evaluated. Westinghouse completed a Safety Evaluation SECL-95-015, 
``OT[DELTA]T and OP[DELTA]T Reactor Trip Response Time Safety 
Evaluation'' to support the revision to the time requirements. The 
evaluation concludes that, as long as the total response time 
remains [less than or equal to] 8 seconds, the safety analyses 
acceptance criteria continue to be met. The OT[DELTA]T and 
OP[DELTA]T trip functions are unaffected by the change.
    The following Updated Final Safety Analysis Report (UFSAR) 
Chapter 15 events trip on OT[DELTA]T: Loss of Electric Load/Turbine 
Trip, Uncontrolled RCCA Bank Withdrawal at Power, CVCS Malfunction 
that Results in a Decrease in the Boron Concentration in the Reactor 
Coolant, and Inadvertent Opening of a Pressurizer Safety or Relief 
Valve. In addition, the following events trip on OP[DELTA]T: 
Steamline Break at Hot Full Power for Core Response, and Steamline 
Break Superheat Analysis. These events have been reviewed for a 
change in the distribution of time responses for OT[DELTA]T and 
OP[DELTA]T. The review concludes that the time response 
redistribution did not result in a minimum DNBR lower than the 
safety analyses limit, did not result in a fuel centerline melt, nor 
did the superheated steam releases change from those currently 
existing. Therefore, the radiological consequences for these events 
do not increase as a result of the less restrictive time response 
breakdown. Thus, the proposed amendment does not result in an 
increase in the probability or consequences of a previously 
evaluated accident.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The OT[DELTA]T and OP[DELTA]T trip functions are unaffected by 
the change. Electronic filtering of the RTD signal has been 
included, changing the dynamic compensation term of OT[DELTA]T and 
OP[DELTA]T setpoint equations. No other changes to the setpoint 
equation result from the proposed modification.
    The added 7300 hardware is compatible with the existing 7300 
electronic hardware now used. All changes to the 7300 protection 
cabinets have been qualified. The proposed system is functionally 
equivalent to the existing one. The proposed modification has been 
reviewed for conformance with the Institute of Electrical and 
Electronics Engineers (IEEE) 279-1971 criteria, associated General 
Design Criteria, Regulatory Guides, and other applicable industry 
standards. The single failure criterion is satisfied by the proposed 
modification, since the independence of redundant protection sets is 
maintained. The new RTD/thermowell system meets the equipment 
seismic and environmental qualification requirements of IEEE 
standards 344-1975 and 323-1974, respectively. The proposed changes 
do not affect the protection system capabilities to initiate a 
reactor trip. The 2 of 4 voting coincidence logic of the protection 
sets is maintained. Therefore, the proposed modification meets all 
appropriate IEEE criteria, industry standards and other guidelines.
    In addition, the RTD outputs are used for rod control, turbine 
runback, pressurizer level and other control systems. These control 
systems receive the signal after it has been processed at the 7300 
cabinets and are therefore unaffected by the proposed modification.
    The design and installation of the thermowells is in accordance 
with the American Society of Mechanical Engineers (ASME) Code 
requirements. However, should a thermowell fail at the RCS pressure 
boundary, the resulting accident is enveloped by current design 
basis accident analyses. Thus, implementation of the proposed 
amendment does not create the possibility of a new or different kind 
of accident from any of those previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The 7300 protection cabinets calculate individual loop [DELTA]T 
and Tavg, based on the output of the RTDs. These values are 
used in the OT[DELTA]T and OP[DELTA]T reactor protection trip 
signals. Electronic filtering of the RTD signal will be included, 
changing the dynamic compensation term of OT[DELTA]T and OP[DELTA]T 
setpoint equations. No other changes to the setpoint equation result 
from the proposed modification. Although the total response time 
used as input into the safety analyses is unaffected by the proposed 
modification, the distribution of response times between the total 
lag (pipe transport delay, RTD response and electronic filter delay) 
and the electronic delay has changed. The UFSAR events which rely on 
OT[DELTA]T and OP[DELTA]T trips have been evaluated. The evaluation 
concludes that the safety analyses acceptance criteria continue to 
be met, since the total response time is consistent with the safety 
analyses. The OT[DELTA]T and OP[DELTA]T trips function in the same 
manner to terminate DNB-related transients. The reliability of the 
reactor protection system is unaffected by this change. Thus, the 
proposed modification does not involve a significant reduction in 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    NRC Project Director: Robert A. Capra.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 
and 2, Will County, Illinois

    Date of amendment request: May 17, 1995.
    Description of amendment request: The proposed amendment would 
modify the technical specifications to allow steam generator tubes to 
be repaired using the tungsten inert gas (TIG) welded sleeve process 
developed by ABB Combustion Engineering (ABB/CE), remove the ability to 
repair steam generator tubes using the Babcock & Wilcox Nuclear 
Technologies (BWNT) kinetically welded sleeve process, and increase the 
requirement to inspect the number of sleeved tubes from 3 percent of 
the total number of sleeved tubes in all four steam generators (SGs) or 
all sleeved tubes in one steam generator to 20 percent of each sleeve 
design installed. The proposed amendments would also delete the 
requirement to conduct additional corrosion testing to establish the 
design life for the BWNT kinetically welded sleeve in the presence of a 
crevice.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or 

[[Page 35065]]
consequences of an accident previously evaluated.
    The proposed amendment allows the ABB/CE TIG welded tubesheet 
sleeves and tube support plate sleeves to be used as an alternate 
tube repair method for Byron and Braidwood Units 1 and 2 Steam 
Generators (SGs). The sleeve configuration was designed and analyzed 
in accordance with the criteria of Regulatory Guide (RG) 1.121 and 
Section III of the ASME Code. Fatigue and stress analyses of the 
sleeved tube assemblies produce acceptable results for both types of 
sleeves as documented in ABB/CE Licensing Report CEN-621-P, Revision 
00, ``Commonwealth Edison Byron and Braidwood Unit 1 & 2 Steam 
Generator Tube Repair Using Leak Tight Sleeves, FINAL REPORT,'' 
April 1995. Mechanical testing has shown that the structural 
strength of the sleeves under normal, faulted, and upset conditions 
is within the acceptable limits specified in RG 1.121. Leakage rate 
testing for the tube sleeves has demonstrated that primary to 
secondary leakage is not expected during any plant condition. The 
consequences of leakage through the sleeved region of the tube is 
fully bounded by the existing steam generator tube rupture (SGTR) 
analysis included in the Byron and Braidwood Updated Final Safety 
Analysis Report (UFSAR).
    The current Technical Specification 3.4.6.2.c primary to 
secondary leakage limit of 150 gallons per day (gpd) through any one 
SG ensures that SG tube integrity is maintained in the event of main 
steam line break (MSLB) or loss of coolant accident (LOCA). The RG 
1.121 criteria for establishing operational leakage rate limits 
require a plant shutdown based upon a leak-before-break 
consideration to detect a free span crack before a potential tube 
rupture. The 150 gpd limit will continue to allow for early leakage 
detection and require a plant shutdown in the event of the 
occurrence of an unexpected crack resulting in leakage that exceeds 
the TS limit.
    The sleeves are designed to allow inservice inspection of the 
pressure retaining portions of the sleeve and parent tube. Inservice 
inspection is performed on all sleeves following installation to 
ensure that each sleeve has been properly installed and is 
structurally sound. Periodic inspections are performed in subsequent 
refuel outages to monitor sleeve degradation on a sample basis. The 
eddy current technique used for inspection will be capable of 
detecting both axial and circumferential flaws. A 20% sample of the 
sleeves are inspected each refuel outage. In the event that an 
imperfection exceeding the repair limit is detected an additional 
20% sample will be inspected. The inspection scope is expanded to 
100% of the sleeves should a repairable defect be found in the 
second sample. Tubes that contain defects in a sleeve, which exceed 
the repair limit, will be removed from service. This ensures that 
sleeve and tube structural integrity is maintained.
    The proposed TS change to support the installation of TIG welded 
sleeves does not adversely impact any previously evaluated design 
basis accident. The effect of sleeve installation on the performance 
of the SG was analyzed for heat transfer, flow restriction, and 
steam generation capacity. The sleeves reduce the risk of primary to 
secondary leakage in the SG. The installation of ABB/CE sleeve 
results in a hydraulic flow restriction that is dependent on the 
number and types of sleeves installed. The reduction in primary 
system flow rate is a small percentage of the flow rate reduction 
seen from plugging one tube and is a preferable alternative when 
considering core margins based on minimum reactor coolant system 
flow rates. The sleeving installation will result in a resistance to 
primary coolant flow through the tube for other evaluated accidents. 
The results of the analyses and testing, as well as industry 
operating experience, demonstrate that the sleeve assembly is an 
acceptable means of maintaining tube integrity. In summary, 
installation of sleeves does not substantially affect the primary 
system flow rate or the heat transfer capability of the steam 
generators.
    The sleeve sample size has been increased from 3% of the sleeved 
tubes in all four steam generators to include an eddy current 
inspection of a minimum of 20% of each sleeve design installed. 
Increasing the sample size of the sleeves to be inspected will 
increase the monitoring of tubes using sleeves for any further 
degradation while they remain in service. If the sample identifies a 
sleeve with an imperfection of greater than the repair limit, an 
additional 20% of the sleeves shall be inspected. The sleeves that 
have identified imperfections of greater than the repair limit shall 
be removed from service. Increasing the monitoring of the sleeves 
will assist in the early detection of a tube or sleeve imperfection 
and limit the probability of occurrence of an accident previously 
evaluated in the UFSAR.
    Installation of the sleeves can be used to repair degraded tubes 
by returning the condition of the tubes to their original design 
basis condition for tube integrity and leak tightness during all 
plant conditions. The tube bundle overall structural and leakage 
integrity will be increased with the installation of the sleeves 
reducing the risk of primary to secondary leakage in the SG while 
maintaining acceptable reactor coolant system flow rates. Therefore 
sleeving will not increase the probability of occurrence of an 
accident previously evaluated.
    Removal of the BWNT kinetically welded sleeve process as an 
approved SG tube repair methodology and not completing the 
additional corrosion testing necessary to establish the design life 
for the BWNT kinetically welded sleeve in the presence of a crevice 
will have no affect on plant operations. There are currently no BWNT 
kinetically welded sleeves installed in the Byron or Braidwood SGs. 
Had there been, plant operations would have still been bounded by 
the existing SGTR analysis in the Byron and Braidwood UFSAR.
    Therefore, these proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The implementation of the proposed sleeving process will not 
introduce significant or adverse changes to the plant design basis. 
Stress and fatigue analyses of the repair has shown the ASME Code 
and RG 1.121 allowable values are met. Implementation of TIG welded 
sleeving maintains overall tube bundle structural and leakage 
integrity at a level consistent with that of the originally supplied 
tubing. Leak and mechanical testing of the sleeves support the 
conclusions that the sleeve retains both structural and leakage 
integrity during all conditions. Repair of a tube with a sleeve does 
not provide a mechanism that result in an accident outside of the 
area affected by the sleeve.
    Any hypothetical accident as a result of potential tube or 
sleeve degradation in the repaired portion of the tube is bounded by 
the existing SGTR analysis. The SGTR analysis accounts for the 
installation of sleeves and the impact on current plugging level 
analyses. The sleeve design does not affect any other component or 
location of the tube outside of the immediate area repaired.
    The current Technical Specification 3.4.6.2.c primary to 
secondary leakage limit of 150 gpd through any one SG ensures that 
SG tube integrity is maintained in the event of an MSLB or LOCA. The 
limit will provide for leakage detection and a plant shutdown in the 
event of the occurrence of an unexpected single crack resulting in 
excessive tube leakage. The leakage limit also provides for early 
detection and a plant shutdown prior to a postulated crack reaching 
critical crack lengths for MSLB conditions.
    Inservice inspections are performed following sleeve 
installation to ensure proper weld fusion has occurred to maintain 
structural integrity. The post installation inspection also serves 
as baseline data to be used for comparison during future 
inspections. Periodic eddy current inspections monitor the pressure 
retaining portions of the sleeve and parent tube for degradation. 
Eddy current techniques will be employed that are sensitive to axial 
and circumferential degradation.
    Increasing the sample size of tubes repaired using either 
sleeving process during each scheduled inservice inspection will 
increase the monitoring of these tubes for any further degradation. 
The improved monitoring and evaluation of the tube and the sleeves 
assures tube structural integrity is maintained or the tube is 
removed for service.
    Corrosion testing of typical sleeve-tube configurations was 
performed to evaluate local stresses, sleeve life, and resistance to 
primary and secondary side corrosion. The tests were performed on 
stress relieved and as-welded (non-stress relieved) sleeve-tube 
joints. Using the corrosion test data in conjunction with finite 
element analyses of the local stress, the stress relieved joint life 
was determined to be in excess of 40 years. The ABB/CE TIG welded 
sleeve operating experience in the industry has shown no sleeve 
failures due to service induced degradation in sleeves that were 
installed with acceptable inspection results. This experience 
includes the stress relieved and 

[[Page 35066]]
as-welded sleeve configurations. ComEd will stress relieve all sleeves 
at Byron and Braidwood as specified in the Technical Report.
    Removal of the BWNT kinetically welded sleeve process as an 
approved SG tube repair methodology and not completing the 
additional corrosion testing necessary to establish the design life 
for the BWNT kinetically welded sleeve in the presence of a crevice 
will not create the possibility of a new or different type of 
accident from any accident previously evaluated. Repair of an SG 
tube with a BWNT kinetically welded sleeve would not have provided a 
mechanism that resulted in an accident outside of the area affected 
by the sleeve. Any hypothetical accident as a result of potential 
tube or sleeve degradation in the repaired portion of the tube would 
have been bounded by the existing SGTR analysis. The SGTR analysis 
accounts for the installation of sleeves and the impact on current 
plugging level analyses. The sleeve design does not affect any other 
component or location of the tube outside of the immediate area 
repaired. Furthermore, there are currently no BWNT kinetically 
welded sleeves installed in the Byron or Braidwood SGs.
    Therefore, the proposed changes do not create the possibility of 
a new or different type of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The TIG welded sleeving repair of degraded steam generator tubes 
has been shown by analysis to restore the integrity of the tube 
bundle to its original design basis condition. The safety factors 
used in the design of the sleeves for the repair of degraded tubes 
are consistent with the safety factors in the ASME Boiler and 
Pressure Vessel Code used in steam generator design. The design of 
the ABB/CE SG sleeves has been verified by testing to preclude 
leakage during normal and postulated accident conditions.
    The portions of the installed sleeve assembly which represents 
the reactor coolant pressure boundary can be monitored for the 
initiation and progression of sleeve/tube wall degradation, thus 
satisfying the requirement of RG 1.83. The portion of the SG tube 
bridged by the sleeve joints is effectively removed from the 
pressure boundary, and the sleeve then forms the new pressure 
boundary. The sleeve enhances the safety of the plant by 
reestablishing the protective boundaries of the steam generator. 
Keeping the tube in service with the use of a sleeve instead of 
plugging the tube and removing it from service increases the heat 
transfer efficiency of the steam generator. During each scheduled 
inservice inspection, each sleeve inspected and found to have 
unacceptable degradation shall be removed from service. The effect 
on the design transients and the accident analyses have been 
reviewed based on the installation of sleeves equal to the tube 
plugging level coincident with the minimum reactor coolant flow 
rate. Evaluation of the installation of sleeves was based on the 
determination that LOCA evaluations for the licensed minimum reactor 
coolant flow bound the combined effect of tube plugging and sleeving 
up to an equivalent of the actual plugging limit. Sleeving results 
in a fractional amount of the plugging limitation of one tube and is 
a preferable alternative when considering core margins based on 
minimum reactor coolant system flow rates. The sleeving installation 
will result in a resistance to primary coolant flow through the 
tube. The primary coolant flow through the ruptured tube is reduced 
by the influence of the installed sleeve, thereby reducing the 
consequences to the public due to a SGTR event.
    A SG sleeve removes an indication of a possible leak source from 
the reactor coolant system (RCS) pressure boundary, eliminating the 
potential of a primary-to-secondary leak. The structural integrity 
of the tube is maintained by the sleeve and sleeve-to-tube joint.
    Installation of either tube sheet or tube support plate sleeves 
will increase the protective boundaries of the steam generators and 
will not reduce the margin of safety.
    Removal of the BWNT kinetically welded sleeve process as an 
approved SG tube repair methodology and not completing the 
additional corrosion testing necessary to establish the design life 
for the BWNT kinetically welded sleeve in the presence of a crevice 
will not result in a reduction in the margin of safety. There are 
currently no BWNT kinetically welded sleeves installed in the Byron 
or Braidwood SGs. SG tube integrity will be maintained by applying 
an alternate NRC approved repair methodology or removing the SG tube 
from service by plugging.
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    NRC Project Director: Robert A. Capra.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of amendment request: April 11, 1995.
    Description of amendment request: The proposed amendments would 
allow a one-time extension of specific LaSalle, Units 1 and 2, 18 month 
Technical Specification Surveillance Requirements to allow surveillance 
testing to coincide with the LaSalle, Unit 1, seventh refueling outage 
(L1R07). The shutdown for L1R07 has been rescheduled from September 
1995 until early 1996. The proposed extensions apply to: Calibrations 
and functional testing of isolation actuation instrumentation, 
emergency core cooling system actuation instrumentation, and 
recirculation pump trip actuation instrumentation; leakage testing of 
reactor coolant system isolation valves; inspection of fire rated 
seals; functional testing of mechanical snubbers; inspections of 
emergency diesel generators; and testing of batteries, battery 
chargers, and other electrical components.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated because:
    The proposed change is temporary and allows a one-time extension 
of specific surveillance requirements for Unit 1 Cycle 7 to allow 
surveillance testing to coincide with the seventh refueling outage. 
The proposed surveillance interval extension is short and will not 
cause a significant reduction in system reliability nor affect the 
ability of the systems to perform their design function. Current 
monitoring of plant conditions and continuation of the surveillance 
testing required during normal plant operation will continue to be 
performed to ensure conformance with Technical Specification 
operability requirements. Therefore, this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated because:
    Extending the surveillance interval for the performance of 
specific testing will not create the possibility of any new or 
different kind of accidents. No changes are required to any system 
configurations, plant equipment, or analyses. Therefore, this change 
will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    (3) Involve a significant reduction in the margin of safety 
because:
    Surveillance interval extensions will not impact any plant 
safety analyses since the assumptions used will remain unchanged. 
The safety limits assumed in the accident analyses and the design 
function of the equipment required to mitigate the consequences of 
any postulated accidents will not be changed since only the 
surveillance test interval is being extended. Historical performance 
generally indicates a high degree of reliability, and surveillance 

[[Page 35067]]
testing performed during normal plant operation will continue to be 
performed to verify continued Operability of affected systems, 
structures and components. Therefore, the plant will be maintained 
within the analyzed limits, and the proposed extension will not 
significantly reduce the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    NRC Project Director: Robert A. Capra.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of amendment request: May 19, 1995.
    Description of amendment request: The proposed amendments would 
revise the technical specification requirement to verify each fire 
protection valve is in the correct position at least once per 31 days. 
The proposed change will retain a monthly visual inspection of the fire 
protection valves that are accessible during plant operation. However, 
the interval for visual surveillance of those valves considered not 
accessible during plant operation will be changed to at least once per 
18 months.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated because: The 
proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated in 
the UFSAR [Updated Final Safety Analysis Report]. The proposed 
change only changes the testing frequency for valves that are 
inaccessible during power operation. A check of the LaSalle LER 
database for the entire operating lifetime of LaSalle Units 1 and 2 
was performed, and there has not been any instances in which any 
Technical Specification related Fire Protection valves have been 
found out of position. Therefore, the change to the frequency of 
testing will have no affect on the capability of fire suppression 
water systems, since all Technical Specification fire protection 
valves, both accessible and inaccessible at power operation, have a 
plant history of 100% correct valve lineup during monthly 
surveillances. Additionally, all fire protection valves that are in 
the fire suppression water flow path are either locked or seal wired 
in the required position at all times. The change does not impact 
the probability of any fire or other accident occurrence. Therefore, 
the proposed change does not cause an increase in the probability or 
consequences of an accident previously evaluated.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated because:
    The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
in the UFSAR. The proposed change only changes the testing frequency 
for valves that are inaccessible during power operation. The change 
to the frequency of testing will have no effect on the capability of 
fire suppression water systems, since the valves, both accessible 
and inaccessible at power operation, have a plant lifetime history 
of 100% correct valve lineup during monthly surveillances. 
Additionally, these valves are locked or sealed in the required 
position at all times. The change does not alter the performance of 
the fire suppression water system, and therefore introduces no new 
failure modes. With no alteration or degradation to equipment or 
system operation, the change introduces no new accident or 
malfunction.
    (3) Involve a significant reduction in the margin of safety 
because:
    The proposed change does not reduce the margin as defined in the 
bases for any Technical Specification. The proposed change only 
changes the testing frequency for all Technical Specification fire 
protection valves that are inaccessible during power operation. The 
plant history of 100% correct valve lineup for the Technical 
Specification fire protection valves, combined with the fact that 
these valves are always locked or sealed in the required position 
ensures that the bases' minimum OPERABILITY requirements of the fire 
suppression systems are met.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    NRC Project Director: Robert A. Capra.

Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion 
Nuclear Power Station, Units 1 and 2, Lake County, Illinois

    Date of amendment request: May 31, 1995.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications and incorporate new acceptance 
criteria for steam generator tubes with degradation in the tubesheet 
roll expansion region.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability of occurrence or consequences of an accident 
previously evaluated.
    Application of the F* criteria to degraded steam generator tubes 
will not affect any of the initiators or precursors of any accident 
previously evaluated. Application of the proposed change will not 
increase the likelihood that a transient initiating event will occur 
because transients are initiated by equipment malfunction and/or 
catastrophic system failure. The proposed change will allow a new 
criteria to be applied to disposition steam generator tubes that are 
degraded in the tubesheet roll transition region. The F* criteria 
specify a minimum length of tubing which must be free from any 
indication of degradation. Below the F* length, any type or size of 
indication, including complete circumferential through wall 
cracking, will not impact the structural integrity of the tube with 
respect to pull out forces during normal operation or accident 
conditions, and does not significantly affect the leakage behavior 
of the tube. While the Zion UFSAR does not specifically address the 
Feedwater Line Break (FLB) accident, the FLB event was used as the 
limiting event in the evaluation of the F* criteria. The FLB 
pressure differential of 2650 psi maximizes the axial loading on the 
tube for pull out considerations and is bounding. In addition, the 
close proximity of the tubesheet to the tube will prevent tube 
rupture or collapse of the tube in the tubesheet span. Because 
application of the F* criteria will ensure that degraded tubes will 
provide the same structural integrity as an original undegraded tube 
during normal operation and accident and accident conditions, the 
probability of occurrence of an accident previously evaluated is not 
significantly increased.
    Application of the F* criteria will not significantly increase 
the consequences of any accident previously evaluated. The F* 
criteria ensure that sufficient length of undegraded tube exists to 
maintain structural integrity and preclude significant leakage. Due 
to the proximity of the tubesheet to the tube, any leakage from 
degradations below the F* length would be negligible and would be 
well below the Technical Specification limits established for steam 
generator 

[[Page 35068]]
leakage. Tube rupture as a result of indications below the F* distance 
is precluded because the tubesheet prevents outward expansion of the 
tube in response to internal pressure.
    The relationship between the tubesheet region leak rate at the 
most limiting postulated accident conditions relative to that for 
normal plant operating conditions has been assessed. For the 
postulated leak source within the roll expansion, increasing the 
differential pressure on the tube on the tube wall increases the 
driving head for the leak; however, it also increases the tube to 
tubesheet loading.
    For a leak source below the F* Distance, the maximum assumed 
pressure differential results in an insignificant leak rate relative 
to that which could be associated with normal plant operation. This 
is a result of the increased tube to tubesheet loading associated 
with the increased differential pressure. Thus for a circumferential 
indication within the roll expansion that is left in service in 
accordance with F* criteria, any leakage under accident conditions 
would be less than that experienced under normal operating 
conditions. Therefore, any leakage under accident conditions would 
be less than the existing Technical Specification leakage limit, 
which is consistent with accident analysis assumptions. Steam 
generator tube integrity must be maintained under the postulated 
loss of coolant accident condition of secondary-to-primary 
differential pressure. Based on tube collapse strength 
characteristics, the constraint provided to the tube by the 
tubesheet gives a margin between the tube collapse strength and the 
limiting secondary-to-primary differential pressure condition, even 
in the presence of circumferential or axial indications. The maximum 
secondary to primary differential pressure during a postulated LOCA 
is 1005 psi. This value is significantly below the residual preload 
between the tubes and the tube sheet. Therefore, no significant 
secondary to primary leakage would be expected to occur.
    In addition, the proposed changes will not affect the ability to 
safely shut down the operating unit and mitigate the consequences of 
an accident because the proposed changes will not necessitate 
changes to the emergency procedures governing accident conditions or 
plant recovery.
    Administrative and typographical changes are proposed to correct 
previous grammatical errors, to eliminate a parenthetical note that 
could cause confusion when applying the proposed requirements, and 
to make the terminology used in the Bases section consistent with 
the definitions provided in Specification 4.3.1. Those proposed 
changes will not increase the probability of occurrence or 
consequence of any accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes to the Technical Specifications do not 
involve the addition of any new or different types of safety related 
equipment nor do they involve the operation of any equipment 
required for safe operation of the facility in a manner different 
from those addressed in the UFSAR. No safety related equipment or 
function will be altered as a result of the proposed changes. Also, 
the procedures governing normal plant operation and recovery from an 
accident are not changed by the application of the F* criteria. The 
F* criteria will allow the use of an alternate method to plugging or 
sleeving to repair steam generator tubes with degradation in the 
tubesheet region. The F* criteria ensure that both the structural 
integrity and leak tight nature of the steam generator tube will be 
equivalent to the original tube. Since no new failure modes or 
mechanisms are introduced by the proposed changes, no new or 
different type of accident is created.
    Administrative and typographical changes are proposed to correct 
previous grammatical errors, to eliminate a parenthetical note that 
could cause confusion when applying the proposed requirements, and 
to make the terminology used in the Bases section consistent with 
the definitions provided in Specification 4.3.1. Those proposed 
changes will not create the possibility of a new or different kind 
of accident from those previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    Plant safety margins are established through Limiting Conditions 
for Operation (LCOs), limiting safety system settings, and safety 
limits specified in Technical Specifications. There will be no 
changes to the LCOs, limiting safety system settings, or the safety 
limits as a result of the proposed changes. Application of the F* 
criteria will allow degraded steam generator tubes to be repaired by 
an alternative method to plugging or sleeving. Steam generator tube 
plugging decreases the total primary reactor coolant flow rate and 
heat transfer capability of the steam generator. While steam 
generator tube sleeving only slightly reduces the reactor coolant 
flow rate, large numbers of sleeves can have a measurable effect on 
flow rate and can complicate steam generator tube inspection 
activities.
    Application of the F* criteria will allow a repair method that 
will restore the integrity of degraded steam generator tubes and 
will not adversely affect primary system flow rate or heat transfer 
capability. Application of the F* criteria will preserve the heat 
transfer capability of the steam generators and will maintain the 
design margins assumed in the analyses contained in the UFSAR. The 
alternate repair method will also be less complicated, faster, and 
will reduce personnel occupational exposure significantly. Based on 
the above discussion it is concluded that the proposed changes will 
not significantly reduce a margin of safety.
    Administrative and typographical changes are proposed to correct 
previous grammatical errors, to eliminate a parenthetical note that 
could cause confusion when applying the proposed requirements, and 
to make the terminology used in the Bases section consistent with 
the definitions provided in Specification 4.3.1. Those proposed 
changes will not impact any margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Waukegan Public Library, 128 
N. County Street, Waukegan, Illinois 60085.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    NRC Project Director: Robert A. Capra.

Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas 
Nuclear One, Unit Nos. 1 and 2 (ANO-1&2), Pope County, Arkansas

    Date of amendment request: April 4, 1995.
    Description of amendment request: The proposed amendments revise 
requirements associated with the ventilation system that services both 
the Unit 1 and Unit 2 control rooms.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

Criterion 1--Does Not Involve a Significant Increase in the Probability 
or Consequences of an Accident Previously Evaluated.

    The control room emergency ventilation and air conditioning 
systems are not initiators of an accident previously evaluated. 
Extension of the allowable outage time for one inoperable control 
room emergency air conditioning system from 7 days to 30 days is 
acceptable based on the low probability of an event occurring that 
would require control room isolation and a concurrent or subsequent 
failure of the remaining operable control room emergency air 
conditioning system. An evaluation using probabilistic safety 
assessment techniques has shown the frequency of this event to be at 
an acceptably low level (4.67E-6/yr). The ANO-1 surveillance 
requirements for the control room emergency ventilation and air 
conditioning system has been updated for consistency with the ANO-2 
requirements and are consistent with RG 1.52, March 1978, Revision 
2. The relaxation in the ANO-2 Mode of Applicability for the control 
room radiation monitoring instrumentation is acceptable based on the 
fuel handling accident analysis dose consequences. The analysis 
assumes that the control room emergency ventilation system is 
actuated during a fuel handling accident in the containment 
building. This analysis also shows that the dose consequences to the 
control room operators are acceptable in the event of a fuel 
handling analysis in the 

[[Page 35069]]
auxiliary building, assuming that the normal control room ventilation 
system only is in operation. When the unit is in Mode 5 or Mode 6 
(with no handling of irradiated fuel in the containment building), 
no accident condition has been identified that would require the 
control room emergency ventilation system to actuate due to high 
radiation. The remainder of the changes have been made for 
consistency between the ANO-1 and ANO-2 TS and are considered to be 
administrative in nature.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.

Criterion 2--Does Not Create the Possibility of a New or Different Kind 
of Accident from any Previously Evaluated

    The control room emergency ventilation and air conditioning 
systems are not accident initiators. The proposed changes introduce 
no new mode of plant operation and no new possibility for an 
accident is introduced by modifying the ANO-1 surveillance testing 
requirements for the control room emergency ventilation and air 
conditioning systems.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

Criterion 3--Does Not Involve a Significant Reduction in the Margin of 
Safety

    With the exception of the AOT extension and the relaxation of 
the ANO-2 Mode of Applicability for the control room radiation 
monitoring instrumentation, all the ANO-1 and ANO-2 changes are 
considered administrative or more restrictive and are intended to 
clarify and make consistent the requirements of the control room 
emergency habitability equipment. Although the AOT extension does 
involve an incremental reduction in the margin of safety due to a 
slight increase in the frequency of an event requiring control room 
isolation, followed by failure of the operable emergency control 
room chiller, a probabilistic safety assessment has shown this 
slight increase in frequency (approximately 3.58E-6/yr) to be 
acceptably low. The relaxation in the ANO-2 Mode of Applicability 
for the control room radiation monitoring instrumentation is 
acceptable based on the fuel handling accident analysis dose 
consequences. The analysis assumes that the control room emergency 
ventilation system is actuated during a fuel handling accident in 
the containment building. This analysis also shows that the dose 
consequences to the control room operators are acceptable in the 
event of a fuel handling analysis [sic., accident] in the auxiliary 
building, assuming that the normal control room ventilation system 
only is in operation. When the unit is in Mode 5 or Mode 6 (with no 
handling of irradiated fuel in the containment building), no 
accident condition has been identified that would require the 
control room emergency ventilation system to actuate due to high 
radiation.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502.
    NRC Project Director: William D. Beckner.

Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas 
Nuclear One, Unit Nos. 1 and 2 (ANO-1&2), Pope County, Arkansas

    Date of amendment request: April 4, 1995.
    Description of amendment request: The proposed amendments delete 
requirements to perform inservice inspections of reactor coolant pump 
flywheels at both Unit 1 and Unit 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

Criterion 1--Does Not Involve a Significant Increase in the Probability 
or Consequences of an Accident Previously Evaluated.

    Missile generation from a reactor coolant pump (RCP) flywheel 
could damage the reactor coolant system, the containment, or other 
equipment or systems important to safety. The fracture mechanics 
analyses conducted to support the change shows that a preexisting 
crack sized just below detection level will not grow to the flaw 
size necessary to create flywheel missiles within the life of the 
plant. This analysis conservatively assumes minimum material 
properties, maximum flywheel accident speed, location of the flaw in 
the highest stress area and a number of startup/shutdown cycles 
eight times greater than expected. Since an existing flaw in the 
flywheel will not grow to the allowable flaw size under normal 
operating conditions or to the critical flaw size under LOCA 
conditions over the life of the plant, elimination of inservice 
inspections for such cracks during the plant's life will not involve 
a significant increase in the probability of an accident previously 
considered.
    The proposed changes do not increase the amount of radioactive 
material available for release or modify any systems used for 
mitigation of such releases during accident conditions. Therefore, 
these changes do not involve a significant increase in the 
consequences of any accident previously evaluated.

Criterion 2--Does Not Create the Possibility of a New or Different Kind 
of Accident from any Previously Evaluated

    The proposed changes will not change the design, configuration, 
or method of operation of the plant and therefore, will not create 
the possibility of a new or different kind of accident from any 
previously evaluated.

Criterion 3--Does Not Involve a Significant Reduction in the Margin of 
Safety

    Significant conservatisms have been used for calculating the 
allowable flaw size, critical flaw size and crack growth rate in the 
RCP flywheels. These include minimum material properties, maximum 
flywheel accident speed, location of the flaw in the highest stress 
area and a number of startup/shutdown cycles eight times greater 
than expected. Since an existing flaw in the flywheel will not grow 
to the allowable flaw size under normal operating conditions or to 
the critical flaw size under LOCA conditions over the life of the 
plant, elimination of inservice inspections for such cracks during 
the plant's life will not involve a significant reduction in the 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502.
    NRC Project Director: William D. Beckner.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2, Pope County, Arkansas

    Date of amendment request: April 4, 1995.
    Description of amendment request: The proposed amendment revises 
surveillance requirements associated with the main turbine steam 
valves.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

Criterion 1--Does Not Involve a Significant Increase in the Probability 
or Consequences of an Accident Previously Evaluated.

    Modifying the surveillance frequency of the main turbine-
generator (MTG) overspeed protection system introduces no new 
failure mechanism for the machine, so the consequences, of a 
postulated MTG overspeed event are no different than those 
previously evaluated. 

[[Page 35070]]

    As explained in NUREG-1366, ``Improvements to Technical 
Specifications Surveillance Requirements,'' the present surveillance 
test frequency requirements were developed for fossil units and 
carried over to nuclear units due to the similarity in design. 
However, the particulate concentration, phosphate chemistry and 
higher steam temperatures present in earlier fossil secondary 
systems, which were major contributing factors to problems 
identified by these tests, are not present in the Arkansas Nuclear 
One-Unit 2 (ANO-2) secondary systems. The operating history of 
turbine valves at ANO-2 is very good, with no failures identified 
during the performance of overspeed protection system surveillance 
testing. Therefore, that change does not involve a significant 
increase in the probability of any accident previously evaluated.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.

Criterion 2--Does Not Create the Possibility of a New or Different Kind 
of Accident from any Previously Evaluated.

    Because the proposed changes do not alter the design, 
configuration, or method of operation of the plant, they do not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
Criterion 3--Does Not Involve a Significant Reduction in the Margin of 
Safety.

    These proposed changes do not alter the acceptance of any 
surveillance requirements, alter any assumptions used in accident 
analysis, change any actuation setpoints, nor allow operations in 
any configuration not previously evaluated. This change in 
surveillance frequency is based on an operating history of the 
turbine overspeed protection system which indicates that reducing 
the test frequency will have no adverse impact on the continued safe 
operation of the unit.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: William D. Beckner.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Nuclear Generating Plant, Unit No. 3, Citrus County, Florida

    Date of amendment request: May 31, 1995.
    Description of amendment request: The proposed amendment would 
revise the the Technical Specifications (TS) for the Crystal River Unit 
3 to facilitate a 24 month operating cycle by changing the surveillance 
interval for appropriate TS surveillance requirements that are 
generally performed during a refueling outage. Additionally, the 
functional description and the ``Allowable Value'' for three Reactor 
Protection System and one Emergency Feedwater Initiation and Control 
System setpoints would be revised. The quantitative limits for 
determining the operational status of the reactor coolant pumps, the 
main feedwater pumps, and the main turbine would be relocated from the 
TS to the Final Safety Analysis Report (FSAR). The surveillance 
associated with the high radiation setpoint for control room isolation 
would also be changed to reflect that the setpoint is an ``approximate 
value'' instead of an ``Allowable value''. The current specified 
surveillance interval for some equipment and systems which were not re-
evaluated or which could not be justified by the evaluation process 
would not be changed.
    Specifically:
    1. TS Surveillance Requirements (SR) 3.3.1.6, SR 3.3.5.3, SR 
3.3.6.1, SR 3.3.9.2, SR 3.3.10.2, SR 3.3.11.3, SR 3.3.17.2, SR 
3.3.18.2, and SR 3.9.2.2 would be revised to extend the surveillance 
frequency from 18 to 24 months. Also, in TS SR 3.3.17.2 a note would be 
added indicating the frequency for Function 12 is 18 months.
    2. In TS Table 3.3.1-1,
    (a) the Function for ``Reactor Coolant Pump Power Monitor (RCPPM)'' 
would be changed to ``Reactor Coolant Pumps,'' and the ``Allowable 
Value'' column for this function would be revised to delete the 
quantitative value and to indicate ``More than one pump tripped'',
    (b) the Function for ``Main Turbine Trip (Control Oil Pressure)'' 
would be changed to ``Main Turbine,'' and the Allowable Value is 
changed to ``Turbine Tripped'' and
    (c) the Function for ``Loss of Both Main Feedwater Pumps (Control 
Oil Pressure)'' would be changed to ``Main Feedwater Pumps,'' and the 
Allowable Value is changed to ``Both Pumps Tripped''
    3. In TS Table 3.3.11-1, Function 1.a would be changed from ``EFW 
Initiation--Loss of MFW Pumps (Control Oil Pressure)'' to ``EFW 
Initiation--Main Feedwater Pumps,'' and the Allowable Value is changed 
to ``Both Pumps Tripped.''
    4. In TS SR 3.3.16.3, the CHANNEL CALIBRATION setpoint would be 
changed from an allowable value to an approximate setpoint.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence or the consequences of an accident 
previously evaluated. The proposed amendment extends the interval 
between successive refueling outage based surveillances to once 
every 24 months for those surveillances evaluated herein and, 
maintains the existing surveillance interval restriction for those 
systems and equipment not evaluated for extension. The reliability 
of systems and components relied upon to prevent or mitigate the 
consequences of accidents previously evaluated is not degraded 
beyond that obtained from the currently defined refueling outage 
interval. Assurance of system and equipment availability is 
maintained. This change does not involve any change to system or 
equipment configuration. Therefore, this change does not increase 
the probability of occurrence or the consequences of an accident 
previously evaluated.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated. The 
proposed amendment extends the interval between successive refueling 
outage based surveillances to once every 24 months for those 
surveillances evaluated herein and maintains the existing 
surveillance interval restriction for those systems and equipment 
not evaluated for extension. This change does not involve any change 
to system or equipment configuration. Therefore, this change is 
unrelated to the possibility of creating a new or different kind of 
accident from any previously evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety. The proposed amendment extends the interval between 
successive refueling outage based surveillances to once every 24 
months for the surveillances evaluated herein, and maintains the 
existing surveillance interval restriction for those systems and 
equipment not evaluated for extension. The reliability of systems 
and components is not degraded beyond that obtained from the 
currently defined refueling outage interval. Assurance of system and 
equipment availability is maintained.
    Therefore, it is concluded that operation of the facility in 
accordance with the proposed amendment does not involve a 
significant reduction in a margin of safety. The proposed extension 
of the refueling outage interval surveillances to once every 24 
months does not degrade the reliability of systems and components 
beyond that obtained from the currently defined refueling outage 
interval. 

[[Page 35071]]
Reliable performance of the systems and equipment effected by this 
change has been demonstrated.
    Implementation of the proposed amendment will maintain the 
required level of assurance of system and equipment availability. 
The surveillance interval for systems and equipment that have not 
been evaluated for extension are excluded from this request. Thus, 
operation of the facility in accordance with the proposed amendment 
involves no significant hazards considerations.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 32629.
    Attorney for licensee: A.H. Stephens, General Counsel, Florida 
Power Corporation, MAC-A5D, P. O. Box 14042, St. Petersburg, Florida 
33733.
    NRC Project Director: David B. Matthews.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Nuclear Generating Plant, Unit No. 3, Citrus County, Florida.

    Date of amendment request: May 31, 1995.
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TS) for the Crystal River Nuclear 
Plant Unit 3 (CR3) relating to the Once Through Steam Generator's 
(OTSG's) tube inspection acceptance criteria. Currently, the TS specify 
repair limit for removing steam generator tubes from service based on a 
structural evaluation of a simplified model of tubes with uniform 
through wall (T/W) thinning. A recent tube-pull examination at CR3 
identified a number of low signal-to-noise (S/N) tube eddy current 
indications. The licensee indicated that these S/N indications are a 
substantially different morphology from the model used to develop the 
current TS inspection and acceptance limit. As a result of the small 
signal amplitude associated with these S/N indications, they cannot be 
accurately sized by conventional bobbin coil phase angle. Therefore, 
the licensee proposed an alternate methodology for dispositioning the 
S/N indications. The proposed criteria would address both wear and 
Inter-Granular-Attack (IGA) degradation mechanisms. Crack-like eddy 
current indications are not included within the proposed scope.
    Specifically, the licensee proposed to:
    A. Revise TS 5.6.2.10.2, page 5.0-14, ``The results of each sample 
inspection shall be classified into one of the following three 
categories:'' to read: ``The results of each bobbin coil sample 
inspection shall be classified into one of the following three 
categories:''
    B. Revise the Note in TS 5.6.2.10.2, page 5.0-14, ``In all 
inspections, previously degraded tubes whose degradation has not been 
spanned by a sleeve must exhibit a significant increase in the 
applicable imperfection size measurement (> +0.5V bobbin coil amplitude 
increase for S/N indications or >10% further wall penetration for all 
other imperfections) to be included in the below percentage 
calculations.''
    C. Revise the sentence in TS 5.6.2.10.4.a.2, page 5.0-16, ``Eddy-
current* * *as imperfections'' to read: S/N indications with a bobbin 
coil amplitude < 0.9V are considered imperfections. Other eddy current 
testing indications below 20% of the nominal tube wall thickness, if 
detectable, may also be considered as imperfections.
    D. Revise TS 5.6.2.10.4.a.4, page 5.0-16, to read:
    ``Degraded Tube means a tube containing a S/N indication with a 
bobbin coil amplitude  0.9V or other imperfection 
 20% of the nominal wall thickness caused by degradation 
except where all such degradation has been spanned by the installation 
of a sleeve.''
    E. Add TS 5.6.2.10.4.a.7 ``Signal-to-Noise (S/N) indication means 
an indication whose associated bobbin coil amplitude is < 5 times the 
background noise, excluding indications located in the tube sheet 
regions or indications determined to be other than a volumetric 
morphology.''
    F. Renumber 5.6.2.10.4.a.7 to 5.6.2.10.4.a.8, and revise to read: 
Plugging/Sleeving Limit means the imperfection depth at or beyond which 
the tube shall be restored to serviceability by the installation of a 
sleeve or removed from service because it may become unserviceable 
prior to the next inspection. The Limit for S/N indications is equal to 
a bobbin coil amplitude of 2.5V, an axial extent of 0.33 inches, or a 
circumferential extent of 0.6 inches. The Limit is equal to 40% of the 
nominal tube or sleeve wall thickness for other imperfections. No more 
than 5000 sleeves may be installed in each OTSG.
    G. Renumber 5.6.2.10.4.a.8, and 9 to 5.6.2.10.4.a.9 and 10.
    H. Revise TS 5.7.2.c.2, page 5.0-29, to read:
    Location, bobbin coil amplitude, and axial and circumferential 
extent (if determined) for each S/N indication and the location and 
percent of wall thickness penetration for each other indication of an 
imperfection, and
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change will not significantly increase the 
probability or consequences of an accident previously evaluated. The 
relevant accidents are excessive leakage or steam generator tube 
rupture (as a consequence of MSLB [Main steam Line Break] or 
otherwise).
    RG [Regulatory Guide] 1.121 establishes a standard method for 
demonstrating structural integrity under worse-than-DBE [design 
basis Event] conditions. The existing TS is based on this RG. The S/
N disposition strategy continues to rely on this guidance. Current 
TW sizing techniques would allow defects greater than the current TS 
limit of 40% to remain in service since these techniques do not 
accurately measure percent wall penetration for small volume 
indications. The proposed disposition strategy is based in 
measurable eddy current parameters of voltage, axial extent, and 
circumferential extent shown to provide a higher confidence that 
unacceptable flaws are removed from service. Therefore, the 
probability of a Steam Generator Tube Rupture (SGTR) is not 
increased and may well be decreased by implementation of this S/N 
disposition strategy.
    The probability of OTSG tube leakage during normal operation or 
accident conditions is not adversely affected by the proposed S/N 
disposition strategy. Operating history indicates essentially no 
primary to secondary leakage through the OTSG tubes at CR-3. Growth 
rate studies imply this trend could be expected to continue. 
Therefore, current leakage limits are retained. Small volume 
indications which might leak during worse-case FWLB [Feedwater Line 
Break] conditions are addressed in the RG 1.121 evaluation. The 
disposition strategy ensure these indications are removed from 
service as part of the inservice inspection. Once detected, the 
proposed criteria is at least as effective in determining those 
indications which should be removed from service as are the existing 
TS limits.
    The S/N disposition strategy is an integral part of an overall 
effort to better address these and similar phenomena in OTSGs.
    2. The proposed change will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The key `new or different' accidents addressed in this and 
similar proposals is the potential for MSLB-induced multiple SGTR or 
excessive primary-to-secondary leakage during such events. While 
these events are addressed in CR-3 Emergency Operating Procedures 
(EOPs), they are beyond those licensed for the facility.
    However, as noted above, the probability of MSLB induced 
multiple SGTR is reduced by more effective screening and plugging/

[[Page 35072]]
sleeving criteria. The probability of detection and identification of 
tubes which should be removed from service is maintained or improved 
by the S/N disposition strategy. The likelihood of adverse effects 
from plugging sound tubes is reduced. The operation of the OTSG or 
related structures, systems or components is otherwise unaffected.
    3. The proposed change will not involve a significant reduction 
to any margin of safety.
    The margins of safety defined in RG 1.121, including the 
required pressure used in the structural analysis, are retained. The 
probability of detecting degradation is unchanged since bobbin coil 
methods will continue to be the primary means of initial detection. 
The probability of leakage remains acceptably small. The proposed S/
N disposition strategy is an enhancement to the inservice inspection 
of OTSG tubing that will provide a higher level of confidence that 
tubes exceeding the allowable limits are repaired while sound tubes 
are left in service. Based upon results of the various growth rate 
studies, the probability of an accident at the end of cycle is 
essentially the same as the beginning.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 32629.
    Attorney for licensee: A. H. Stephens, General Counsel, Florida 
Power Corporation, MAC-A5D, P. O. Box 14042, St. Petersburg, Florida 
33733.
    NRC Project Director: David B. Matthews.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Plant Units 3 and 4, Dade County, Florida

    Date of amendment request: June 19, 1995.
    Description of amendment request: The licensee proposes to change 
Turkey Point Units 3 and 4 Technical Specifications (TS) by separation 
of the 24-hour emergency diesel generator (EDG) run and hot restart EDG 
test from the loss-of-offsite-power load acceptance test. The licensee 
revised the original amendment request dated March 30, 1995, by letters 
dated May 5, 1995, and June 19, 1995.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which was previously presented in the Federal Register 
(60 FR 27339, May 23, 1995). The licensee concluded that the proposed 
license amendments' revisions do not alter the original conclusion that 
no significant hazards considerations exist pursuant to 10 CFR 50.92.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request and its revisions involve no significant hazards 
consideration.
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199.
    Attorney for licensee: J.R. Newman, Esquire, Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Project Director: David B. Matthews.

Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric 
Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and 
50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2, Appling County, 
Georgia

    Date of amendment request: January 13, 1995, as supplemented by 
letters dated April 5 and June 20, 1995.
    Description of amendment request: The proposed amendments would 
change the Facility Operating Licenses and their corresponding 
Appendices A which contain the Technical Specifications (TS) to permit 
the implementation of the power uprate program at the Edwin I. Hatch 
Nuclear Plant, Units 1 and 2. The Hatch units are currently licensed 
for operation at 2436 megawatts thermal (MWt). The proposed changes 
would redefine the rated thermal power to 2558 MWt, which represents an 
increase of 5% over the current licensed level in accordance with the 
generic boiling water reactor (BWR) power uprate program established by 
the General Electric Company (GE) and approved by the U.S. Nuclear 
Regulatory Commission (NRC) staff in a letter from W. T. Russell, NRC, 
to P. W. Marriott, GE, dated September 30, 1991. Implementation of the 
proposed power uprate at Plant Hatch will result in an increase of 
steam flow to approximately 106% of the current value but will require 
no changes to the basic fuel design. Implementation of this proposed 
power uprate will require minor modifications, such as resetting the 
safety relief setpoints, as well as the calibration of plant 
instrumentation to reflect the uprated power. Plant operating, 
emergency, and other procedure changes will be made where necessary to 
support uprated operation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Will the changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    A. Rated Thermal Power is increased to 2558 MWt on page 3 of the 
Unit 1 Operating License, page 4 of the Unit 2 Operating License, 
and in Section 1.1 (Definitions) of the Units 1 and 2 Technical 
Specifications.
Evaluation

    The changes in the Operating Licenses and Technical 
Specifications were evaluated and it was determined that the 
probability (frequency of occurrence) of design basis accidents 
occurring is not affected by the increased power level, as the 
regulatory criteria established for plant equipment (e.g., ASME 
Code, IEEE standards, NEMA standards, Regulatory Guide criteria) 
will still be complied with at the uprated power level. Scram 
setpoints (equipment settings that initiate automatic plant 
shutdowns) will be established such that there is no significant 
increase in scram frequency due to uprate. No new challenges to 
safety-related equipment will result from power uprate.
    The changes in consequences of hypothetical accidents which 
would occur from 102% of the uprated power, compared to those 
previously evaluated, are in all cases insignificant, because the 
power uprate accident evaluations will not result in exceeding any 
NRC-approved acceptance limits. Enclosure 4 of Reference 1, General 
Electric Report NEDC-32405P, ``Power Uprate Safety Analysis for 
Edwin I. Hatch Plant Units 1 and 2,'' December 1994, investigated 
the spectrum of hypothetical accidents and transients, and showed 
the plant's current regulatory criteria are satisfied at power 
uprate. For example, in the area of core design, the fuel operating 
limits will still be met at the uprated power level, and fuel reload 
analyses will show plant transients meet the criteria accepted by 
the NRC as specified in NEDO-24011, ``GESTAR II.'' Challenges to 
fuel or emergency core cooling system (ECCS) performance were 
evaluated (Section 4.2 of NEDC-32405P) and shown to still meet the 
criteria of 10 [CFR] 50.46 and Appendix K. Challenges to the 
containment were evaluated (Section 4.1 of NEDC-32405P) and shown to 
still meet 10 CFR 50 Appendix A, Criterion 38, Long Term Cooling, 
and Criterion 50, Containment. Radiological release events were 
evaluated (Section 9.2 of NEDC-32405P) and shown to meet the 
criteria of 10 CFR 100 (Unit 1 FSAR Chapter 14 and Unit 2 FSAR 
Chapter 15).
    The results of the analyses discussed above demonstrate that 
operation at the power uprate level does not significantly increase 
the probability or consequences of an accident previously evaluated.
    B. The surveillance test discharge pressure for the standby 
liquid control pump at 41.2 gpm is increased from 1190 psig to 1201 
psig. This value appears in Surveillance Requirement (SR) 3.1.7.7 
and the 

[[Page 35073]]
corresponding Bases Section B 3.1.7 in the Unit 1 and Unit 2 Technical 
Specifications.

Evaluation

    Power uprate operation will result in a 30 psi increase in 
reactor operating pressure. As will be discussed in these proposed 
changes, several pressure-dependent setpoints (including safety 
relief valve [SRV] setpoints) will be increased to preserve current 
margins. Increasing the pressure 11 psi, at which a 41.2 gpm flow 
rate is developed, assures continued conformance to anticipated 
transient without scram (ATWS) criteria at uprated conditions. The 
surveillance test pressure is based on the maximum pressure for an 
ATWS event during the time period when the standby liquid control 
pump is in operation. Section 6.5 of NEDC-32405P discusses the 
capability of these positive displacement pumps. A small increase in 
the SRV setpoints will have no effect on the rated injection flow to 
the reactor. This change, therefore, will not increase the 
probability or consequences of a previously evaluated accident.
    C. The reactor vessel steam dome high pressure allowable value 
for reactor protection system (RPS) instrumentation is increased 31 
psi, consistent with the nominal pressure increase for power uprate. 
The allowable value appears in Section 3.3.1.1, Table 3.3.1.1-1, 
Function 3, in the Unit 1 and Unit 2 Technical Specifications.

Evaluation

    The reactor vessel steam dome high pressure scram limit is 
increased because the steam dome operating pressure is increased. 
Operating pressure for uprated power is increased to assure that 
satisfactory reactor pressure control is maintained. The operating 
pressure was chosen on the basis of steam line pressure drop 
characteristics and the steam flow capability of the turbine. 
Satisfactory reactor pressure control requires an adequate flow 
margin between the uprated operating condition and the steam flow 
capability of the turbine control valves at their maximum stroke. An 
operating dome pressure of 1035 psig, which is 30 psi higher than 
the current operating dome pressure, is expected. Therefore, the 
high pressure scram is increased approximately the same amount to 
preserve existing margins to reactor trips.
    The high pressure scram terminates a pressurization transient 
not terminated by direct scram or high neutron flux scram. The 
setting is maintained above the nominal reactor vessel operating 
pressure and below the specified analytical trip limit used in the 
safety analyses. The revised high pressure scram setpoint will 
preserve the hierarchy of pressure setpoints. This means that the 
high pressure scram setpoint will remain below the opening setpoint 
of the SRVs. The SRV nominal setpoints are also increased 30 psi, as 
discussed in Item G below. This hierarchy of setpoints provides 
assurance that the probability of opening more than one SRV without 
scram intervention is low.
    Since the scram function and the current margins to trip 
avoidance are maintained with revised setpoints, there is no 
significant increase in the probability or consequences of an 
accident previously evaluated.
    D. The ATWS reactor vessel steam dome high pressure 
recirculation pump trip (RPT) allowable value is raised 80 psi. The 
allowable value appears in Section 3.3.4.2, SR 3.3.4.2.3, in the 
Unit 1 and Unit 2 Technical Specifications.

Evaluation

    The ATWS-RPT high pressure setpoint initiates a trip of the 
recirculation pumps, thereby adding negative reactivity following 
events in which a scram does not (but should) occur. Section 5.1.3.2 
of NEDC-32405P discusses this function in detail.
    The current analytical limit for the ATWS-RPT high pressure trip 
is 1150 psig. This value was increased 30 psi in the power uprate 
ATWS safety evaluations to account for the 30 psi increase in vessel 
operating pressure, SRV setpoints, etc. The current allowable value 
in the Technical Specifications is 1095 psig. This allowable value 
was not set by the current analytical limit, but by the range of the 
installed pressure instruments. As part of the power uprate plant 
changes, these pressure instruments will be replaced to accommodate 
higher pressure, and the allowable value, in conjunction with the 
analytical limit used in the safety analysis, will be increased.
    Sections 5.1 and 9.3 of NEDC-32405P show the system can 
adequately perform its ATWS function with the new setpoint. 
Therefore, the proposed change does not cause a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    E. The low-low set (LLS) SRV arming pressure allowable value is 
increased 31 psi, consistent with the increase in operating pressure 
and high pressure scram allowable value. The LLS arming pressure 
allowable value appears in Section 3.3.6.3, Table 3.3.6.3-1, 
Function 1, in the Unit 1 and Unit 2 Technical Specifications.

Evaluation

    The allowable value for the LLS SRV high pressure arming 
setpoint is increased because the high pressure scram setpoint is 
increased. No changes to the LLS arming logic associated with the 
SRV tailpipe pressure switches and the LLS opening and closing 
pressure setpoints are proposed.
    The LLS relief logic mitigates the postulated containment loads 
of subsequent SRV actuations during small or intermediate loss of 
coolant accidents (LOCAs) by extending the time between actuations. 
The LLS logic requires two separate signals to arm itself for 
operation. Specifically, the LLS logic arms when an SRV opens (i.e., 
tailpipe pressure switch) and reactor pressure concurrently exceeds 
the scram setpoint. To preserve the hierarchy of pressure setpoints, 
the high pressure input to the LLS SRV arming logic has the same 
setpoint as the high pressure scram, thus minimizing the potential 
for a spurious SRV opening through the LLS logic without occurrence 
of a reactor scram.
    Increasing the arming setpoint is consistent with increasing the 
high pressure scram setpoint and will not increase the probability 
or consequences of an accident previously evaluated.
    F. Lower the permissible rod line for single-loop operation 
(SLO) below 45 percent core flow from the 80 percent rod line to the 
76 percent rod line. This Technical Specifications limit appears in 
Section 3.4.1 (Figure 3.4.1-1) and the corresponding Bases Section B 
3.4.1 of the Unit 1 and Unit 2 Technical Specifications.
Evaluation

    During development of the generic power uprate program, GE and 
the NRC agreed to maintain the current exclusion region in the 
power-to-flow map related to thermal-hydraulic stability. The 
current limit for SLO is the 80 percent rod line. Power uprate will 
redefine 100 percent rated power and, therefore, rated rod or flow 
control lines. The 76 percent rod line at uprated conditions closely 
corresponds on an absolute, rather than percentage basis, to the 
existing 80 percent rod line.
    Therefore, this proposed Technical Specifications change ensures 
that power uprate operation will not cause a significant increase in 
the probability or consequences of accident previously evaluated.
    G. The SRV lift setpoints in the Units 1 and 2 Technical 
Specifications SR 3.4.3.1 will be increased 30 psi.

Evaluation

    The SRVs are designed to prevent overpressurization of the 
reactor pressure vessel during abnormal operational transients. The 
SRV lift setpoints are increased to accommodate the increase in 
operating pressure that accompanies power uprate. The increase in 
SRV setpoints ensures that adequate margins are maintained so that 
the increase in dome pressure during normal operation does not 
result in an increase in the number of unnecessary SRV actuations. 
The setpoint increase also maintains the hierarchy of pressure 
setpoints described in these proposed changes. Transient evaluations 
include a +3 percent tolerance to the nominal setpoints. As 
described in Section 3.2 of NEDC-32405P, peak vessel pressure 
increases by 3 percent, but remains well below the 1375 psig ASME 
Code limit.
    Although not credited in the transient analysis, GPC installed a 
pressure transmitter system which can electronically actuate the 
SRVs on high vessel pressure. The nominal trip setpoints for its 
actuation correspond with the nominal mechanical lift setpoints in 
the Technical Specifications. The SRV pressure transmitter system 
nominal setpoints will also be increased 30 psi.
    General Electric generically evaluated the adequacy of BWR SRVs 
to operate at uprated temperatures and pressures. The reactor 
operating pressure and temperature increases of less than 40 psi and 
5 deg.F, respectively, used in that evaluation bound the uprated 
Hatch operating conditions.
    The impact of power uprate on the Hatch containment dynamic 
loads due to SRV discharge has also been evaluated. As discussed in 
Section 4.1.2 of NEDC-32405P, the vent thrust loads with power 
uprate were calculated to be less than the loads used in the 
containment analysis. The effects of power uprate on SRV air-
clearing, the 

[[Page 35074]]
discharge line, the pool pressure boundary, and submerged structure 
drag loads are discussed in Section 4.1.2 of NEDC-32405P which 
concludes that the small increase in the setpoint pressure is well 
within the margin in the SRV loads defined in the Mark I Containment 
Long-Term Program. Therefore, power uprate does not impact the Hatch 
SRV load definitions used in the containment analysis, and no 
significant increase in the probability or consequences of an 
accident previously evaluated is caused by this proposed change.
    H. The Limiting Condition for Operation (LCO) and SRs for the 
maximum reactor steam dome pressure will be increased from 1020 psig 
to 1058 psig. This requirement appears in LCO 3.4.10, SR 3.4.10.1, 
and the corresponding Bases in the Unit 1 and Unit 2 Technical 
Specifications.

Evaluation

    As discussed in the Technical Specifications Bases and NEDC-
32405P, the maximum reactor dome pressure is an initial condition of 
the vessel overpressure protection analysis, which assumes a fast 
isolation of all four main steam lines by the main steam isolation 
valves (MSIVs). The reactor scram signal generated directly by the 
valve closure is assumed defeated for this analysis. Instead, the 
scram signal is generated by high neutron flux. The overpressure 
analysis for power uprate assumed an initial dome pressure of 1058 
psig, which represents an increase of 38 psig. This initial pressure 
was chosen approximately 2 percent above the 1035 psig steam dome 
operating pressure expected for power uprate operation. The analysis 
also included the other changes (including SRV setpoints) discussed 
in these proposed changes. Therefore, there is no significant 
increase in the probability or consequences of an accident 
previously evaluated.
    I. The HPCI and RCIC surveillance test pressures in Units 1 and 
2 Technical Specifications SRs 3.5.1.8 and 3.5.3.3, respectively, 
are increased 38 psi.

Evaluation

    The allowable HPCI and RCIC surveillance test pressure is 
increased to correspond with the increase in normal reactor 
operating pressure and LCO/SR on maximum reactor pressure that 
accompanies power uprate. (As discussed in Item H above, the LCO on 
reactor steam dome pressure is increased 38 psi.) The change is 
needed to ensure that pressure and power reductions are not required 
to perform surveillance testing. The requested changes will allow 
the quarterly demonstration of the HPCI and RCIC systems' capability 
to perform at normal reactor operating pressures, which meets the 
original intent of the Technical Specifications.
    The HPCI and RCIC systems have been evaluated and demonstrated 
to be capable of injecting design flow rate at the higher reactor 
pressure as discussed in Sections 4.2 and 3.8 of NEDC-32405P and in 
Reference 2.
    Therefore, these changes will ensure that power uprate operation 
will not cause a significant increase in the probability or 
consequences of an accident previously evaluated.

J. Bases Changes

    Several changes to the Hatch Units 1 and 2 Technical 
Specifications Bases are proposed for consistency with the power 
uprate safety analyses. These proposed changes are in addition to 
the Bases changes corresponding to proposed changes A through I.
    i. The main steam line flow differential pressure setpoints 
(Bases Section B 3.3.6.1.c) and the HPCI/RCIC high flow differential 
pressure setpoints (Bases Section B 3.3.6.3.a and B 3.3.6.4.a) are 
changed for both units.
    The allowable values (in percent of rated) will not change for 
power uprate operation. However, the actual differential pressure 
will change due to the increase in steam flow and pressure.
    ii. The HPCI and RCIC upper design pressure in Bases Sections B 
3.5.1 and B 3.5.3, respectively, is increased 34 psi for both units
    The Bases changes support the design of these high pressure 
systems to pump rated flow from approximately 150 psig up to a 
pressure associated with the first group of SRV setpoints. This 
proposed design pressure conservatively considers the 30 psi higher 
nominal setpoints and 3 percent setpoint drift. The capability of 
the HPCI and RCIC systems to deliver design flows at these pressures 
is discussed in Reference 2, and was reviewed by GE for the Unit 1 
and Unit 2 systems.
    Note that the upper design pressure for HPCI and RCIC is 
different from the surveillance test pressure for HPCI and RCIC 
discussed previously in item I. The maximum surveillance test 
pressure corresponds to reactor operating pressure, since the 
surveillance test is performed when the unit is operating. The HPCI 
and RCIC upper design pressure reflects the capability to inject 
water to the vessel following a reactor scram and isolation.
    iii. The peak post accident containment pressure (Pa) is 
changed to 49.6 psig (Unit 1) and 45.5 psig (Unit 2). These values 
appear in Bases Sections B 3.6.1.1, B 3.6.1.2, and B 3.6.1.4 in each 
unit's Technical Specifications.
    Section 4.1.1.3 of NEDC-32405P discusses the peak short-term 
containment pressure response which was recalculated for power 
uprate conditions. Containment pressure and temperatures remain 
below design limits and are essentially unchanged.
    iv. The main condenser offgas gross gamma activity rate limit of 
240 mci/second will not be changed for power uprate. A statement 
that the current limit is conservative for power uprate conditions 
was added to Bases Section 3.7.6 for both units.
    The Bases derive the current 240 mci/second limit using a rated 
core thermal power limit of 2436 MWt. A slightly higher limit could 
be justified using the uprated power level. However, adequate margin 
exists with the current limit.
    v. The inservice hydrostatic and leak testing pressures shown in 
Bases Section 3.10.1 are increased 33 psi and 30 psi, respectively. 
This change affects each unit's Bases.
    This change is a direct result of the 30 psi increase in normal 
operating pressure proposed for power uprate. The leakage test is 
normally performed at operating pressure and the hydrostatic test at 
approximately 110 percent of operating pressure.
    The above Bases changes Items i-v have been evaluated and will 
not increase the probability or consequences of an accident 
previously evaluated.
    2. Will the changes create the possibility of a new or different 
kind of accident from any accident previously evaluated?

Evaluation

    The Operating License changes in power level and the associated 
Technical Specifications changes discussed previously will not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated, as summarized below.
    Equipment that could be affected by power uprate was evaluated. 
No new operating mode, safety-related equipment lineup, accident 
scenario, or equipment failure mode were identified. The full 
spectrum of accident considerations defined in RG 1.70 was 
evaluated, and no new or different kind of accident was identified. 
Uprate uses already-developed technology and applies it within the 
capabilities of existing plant equipment in accordance with 
presently existing regulatory criteria to include NRC-approved 
codes, standards, and methods. GE has designed BWRs of higher power 
levels than the uprated power of any of the currently operating BWR 
fleet, and no new power dependent accidents have been identified.
    The Technical Specifications changes required to implement power 
uprate require only minor modifications to the plant's 
configuration. All changes were evaluated and found to be 
acceptable.
    3. Will the changes involve a significant reduction in the 
margin of safety?
    A. Rated Thermal Power is increased to 2558 MWt on page 3 of the 
Unit 1 Operating License, page 4 of the Unit 2 Operating License, 
and in Section 1.1 (Definitions) of the Unit 1 and Unit 2 Technical 
Specifications.

Evaluation

    The events analyzed in the FSAR were re-evaluated to demonstrate 
that power uprate can be implemented without exceeding any 
regulatory limit. Because the applicable safety analysis criteria 
and limits are satisfied for power uprate, the margin of safety 
associated with the safety limits and other limits identified in the 
Technical Specifications will be maintained.
    As discussed in NEDC-32405P, the safety margins prescribed by 
the Code of Federal Regulations are maintained by meeting the 
appropriate regulatory criteria. Similarly, the margins provided by 
the application of the ASME design criteria are maintained. Section 
11.4.2 of NEDC-32405P discusses the effects of power uprate on 
safety margins for the following:
    Fuel thermal limits Design basis accidents and the challenges to 
fuel, containment, and radiological releases. Transient analyses. 
Non-LOCA radiological releases. Environmental consequences.
    These evaluations conclude that applicable safety analysis 
criteria and limits are 

[[Page 35075]]
satisfied, and thus, the margin of safety will not be significantly 
reduced.
    B. The surveillance test discharge pressure for the SLC pump at 
41.2 gpm is increased from 1190 psig to 1201 psig. This value 
appears in SR 3.1.7.7 and corresponding Bases Section B 3.1.7 in the 
Unit 1 and Unit 2 Technical Specifications.

Evaluation

    Power uprate operation will result in a 30 psi increase in 
reactor operating pressure. Several pressure-dependent setpoints 
(including SRV setpoints) will be increased to preserve current 
margins. Increasing the pressure 11 psi, at which a 41.2 gpm flow 
rate is developed, assures continued conformance to ATWS criteria at 
uprated conditions. The surveillance test pressure is based on the 
maximum pressure for an ATWS event during the time period when the 
SLC pump is in operation. Section 6.5 of NEDC-32405P discusses the 
capability of these positive displacement pumps. A small increase in 
the SRV setpoints will have no effect on the rated injection flow to 
the reactor.
    For power uprate, the capability of the SLCS to respond with 
adequate margin to an ATWS event was confirmed. The results are 
reported in Section 9.3.1 of NEDC-32405P. The limiting ATWS event 
was an inadvertent MSIV closure. The event was reanalyzed at uprate 
conditions with the higher SRV setpoints and ATWS-RPT setpoints. 
Peak vessel pressure was well below the ASME emergency limit of 1500 
psig. The effect of power uprate on peak clad temperature and 
maximum suppression pool temperature was judged to be negligible, 
because the calculations showed no increase in fuel surface heat 
flux or integrated SRV flow.
    In summary, all ATWS criteria are satisfied and the SLC pumps 
are capable of injecting the required amounts of sodium pentaborate 
at uprated conditions. Therefore, there is no significant decrease 
in the margin of safety.
    C. The reactor vessel steam dome high pressure allowable value 
for RPS instrumentation is increased 31 psi, consistent with the 
nominal pressure increase for power uprate. The allowable value 
appears in Section 3.3.1.1, Table 3.3.1.1-1, Function 3, in the Unit 
1 and Unit 2 Technical Specifications.

Evaluation

    The reactor vessel steam dome high pressure scram limit is 
increased because the steam dome operating pressure is increased. 
Operating pressure for uprated power is increased to assure that 
satisfactory reactor pressure control is maintained. The operating 
pressure was chosen on the basis of steam line pressure drop 
characteristics and the steam flow capability of the turbine. 
Satisfactory reactor pressure control requires an adequate flow 
margin between the uprated operating condition and the steam flow 
capability of the turbine control valves at maximum stroke. An 
operating dome pressure of 1035 psig, which is 30 psi higher than 
the current operating dome pressure, is expected. Therefore, the 
high pressure scram is increased approximately the same amount to 
preserve existing margins to reactor trips.
    The increases in the steam dome high pressure scram instrument 
setpoints for uprated power were evaluated by determining whether 
the high pressure scram, which is used as a backup to other scram 
signals, provides adequate overpressure protection. The evaluation 
demonstrates that the backup protection function, with the revised 
setpoints, continues to provide adequate overpressure protection at 
uprated power conditions by meeting the applicable ASME Code 
criteria. Therefore, there is no significant decrease in the margin 
of safety.
    D. The ATWS reactor vessel steam dome high pressure RPT 
allowable value is raised 80 psi. The allowable value appears in 
Section 3.3.4.2, SR 3.3.4.2.3, in the Unit 1 and Unit 2 Technical 
Specifications.

Evaluation

    The ATWS-RPT high pressure setpoint initiates a trip of the 
recirculation pumps, thereby adding negative reactivity following 
events in which a scram does not (but should) occur. Section 5.1.3.2 
of NEDC-32405P discusses this function in detail.
    For power uprate, the capability of the SLCS to respond to a 
postulated ATWS event with adequate margin was confirmed (Section 
9.3.1 of NEDC-32405P). By reducing reactor power until the SLCS can 
inject the required amounts of sodium pentoborate to achieve full 
shutdown, the RPT also reduces suppression pool temperature for 
isolation cases (also shown to be acceptable for power uprate 
conditions in Section 9.3.1 of NEDC-32405P). Therefore, there is no 
significant decrease in a margin of safety.
    E. The LLS SRV arming pressure allowable value is increased 31 
psi, consistent with the increase in operating pressure and high 
pressure scram allowable value. The LLS arming pressure allowable 
value appears in Section 3.3.6.3, Table 3.3.6.3-1, Function 1, in 
the Unit 1 and Unit 2 Technical Specifications.

Evaluation

    The allowable value for the LLS SRV high pressure arming 
setpoint is increased, because the high pressure scram setpoint is 
increased. No changes to the LLS arming logic associated with the 
SRV tailpipe pressure switches, and the LLS opening and closing 
pressure setpoints are proposed.
    Since this proposed change only affects one of two arming 
signals for LLS, the safety analyses are not affected; therefore, 
there is not a significant change in the margin of safety.
    F. Lower the permissible rod line for SLO below 45 percent core 
flow from the 80 percent rod line to the 76 percent rod line. This 
Technical Specifications limit appears in Section 3.4.1 (Figure 
3.4.1-1) and corresponding Bases Section B 3.4.1 of the Unit 1 and 
Unit 2 Technical Specifications.

Evaluation

    This change to the power versus flow map restricted zone is made 
to maintain the same operating constraints and stability margin that 
were established for the current power level. This change avoids any 
increase in the possibility of occurrence or any increase in the 
potential effects of power oscillations. Therefore, there is no 
significant decrease in a margin of safety.
    G. The SRV lift setpoints in Surveillance Requirement 3.4.3.1 
(both units) will be increased 30 psi.

Evaluation

    The SRVs are designed to prevent overpressurization of the 
reactor pressure vessel during abnormal operational transients. The 
SRV lift setpoints are increased to accommodate the increase in 
operating pressure that accompanies power uprate. The increase in 
SRV setpoints ensures that adequate margins are maintained so that 
the increase in dome pressure during normal operation does not 
result in an increase in the number of unnecessary SRV actuations. 
The setpoint increase also maintains the hierarchy of pressure 
setpoints described in these proposed changes. Transient evaluations 
include a + 3 percent tolerance to the nominal setpoints. As 
described in Section 3.2 of NEDC-32405P, peak vessel pressure 
increases by 3 percent but remains well below the 1375 psig ASME 
Code limit. Therefore, there is no significant decrease in the 
margin of safety.
    H. The Limiting Condition for Operation (LCO) and Surveillance 
Requirements for the maximum reactor steam dome pressure will be 
increased from 1020 psig to 1058 psig. This requirement appears in 
LCO 3.4.10, SR 3.4.10.1, and the corresponding Bases in the Unit 1 
and Unit 2 Technical Specifications.

Evaluation

    As discussed in the Technical Specifications Bases and in 
Section 3.2 of NEDC-32405P, the maximum reactor dome pressure is an 
initial condition of the vessel overpressure protection analysis, 
which assumes a fast isolation of all four main steam lines by the 
main steam isolation valves. It is also used as a sensitivity study 
parameter for certain transient and LOCA events.
    With this revised limit, peak vessel pressure remains below ASME 
Code criteria, transient limits are maintained, and LOCA fuel 
performance satisfies the requirements of 10 CFR 50.46 and 10 CFR 
50, Appendix K. Therefore, there is no significant decrease in a 
margin of safety.
    I. The HPCI and RCIC surveillance test pressures in SRs 3.5.1.8 
and 3.5.3.3, respectively, (both units) are increased 38 psi.

Evaluation

    The allowable HPCI and RCIC surveillance test pressure is 
increased to correspond with the increase in normal reactor 
operating pressure and LCO/SR on maximum reactor pressure that 
accompanies power uprate. (As discussed previously, the LCO on 
reactor steam dome pressure is increased 38 psi.)
    The purpose of the HPCI and RCIC surveillance test is to provide 
periodic demonstration of the systems' ability to perform consistent 
with the requirements of the analyses at the higher operating 
pressure associated with power uprate conditions. An evaluation of 
the HPCI and RCIC systems confirmed their ability to operate at 
slightly higher turbine speed and provide design flow 

[[Page 35076]]
at power uprate conditions. System performance will be confirmed during 
the initial power ascension to uprated conditions (and periodically 
thereafter per the Technical Specifications). Therefore, there is no 
significant decrease in the margin of safety.

J. Bases Changes

    Several changes to the Hatch Units 1 and 2 Technical 
Specifications Bases are proposed for consistency with the power 
uprate safety analyses. These proposed changes are in addition to 
the Bases changes corresponding to proposed changes A through I.
    i. The main steam line flow differential pressure setpoints, as 
shown in Bases Section B 3.3.6.1.c, and the HPCI/RCIC high flow 
differential pressure setpoints (Units 1 and 2 Bases Sections B 
3.3.6.3.a and B 3.3.6.4.a) are changed.
    The allowable values (in percent of rated) will not change for 
power uprate operation. However, the actual differential pressure 
will change due to the increase in steam flow and pressure.
    ii. The HPCI and RCIC upper design pressure in Units 1 and 2 
Bases Sections B 3.5.1 and B 3.5.3, respectively, is increased 34 
psi.
    The Bases changes support the design of these high pressure 
systems to pump rated flow from approximately 150 psig up to a 
pressure associated with the first group of SRV setpoints. This 
proposed design pressure conservatively considers the 30 psi higher 
nominal setpoints and 3 percent setpoint drift. The capability of 
the Unit 1 and Unit 2 HPCI and RCIC systems to deliver design flows 
at these pressures was reviewed by GE and is discussed in Reference 
2.
    iii. The peak post accident containment pressure (Pa) is 
changed to 49.6 psig (Unit 1) and 45.5 psig (Unit 2). These values 
appear in Units 1 and 2 Bases Sections B 3.6.1.1, B 3.6.1.2, and B 
3.6.1.4.
    Section 4.1.1.3 of NEDC-32405P discusses the peak short-term 
containment pressure response which was recalculated for power 
uprate conditions. Containment pressure and temperatures remain 
below design limits and are essentially unchanged.
    iv. The main condenser offgas gross gamma activity rate limit of 
240 mci/second will not be changed for power uprate. A statement 
that the current limit is conservative for power uprate conditions 
was added to Units 1 and 2 Bases Section 3.7.6.
    The Bases derive the current 240 mci/second limit using a rated 
core thermal power limit of 2436 MWt. A slightly higher limit could 
be justified using the uprated power level. However, adequate margin 
exists with the current limit.
    v. The inservice hydrostatic and leak testing pressures shown in 
Units 1 and 2 Bases Section 3.10.1 are increased 33 psi and 30 psi, 
respectively.
    This change is a direct result of the 30 psi increase in normal 
operating pressure proposed for power uprate. The leakage test is 
normally performed at operating pressure and the hydrostatic test at 
approximately 110 percent of operating pressure.
    The above Bases changes i-v were evaluated, and there is no 
significant decrease in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 
20037.
    NRC Project Director: Herbert N. Berkow.

Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric 
Authority of Georgia, City of Dalton, Georgia, Docket No. 50-366, Edwin 
I. Hatch Nuclear Plant, Unit 2, Appling County, Georgia

    Date of amendment request: April 14, 1995.
    Description of amendment request: The licensee proposes to revise 
Plant Hatch Unit 2 Technical Specifications (TS) to eliminate selected 
response time testing requirements from the TS. Specifically, the 
response time testing to be eliminated includes sensors and specified 
loop instrumentation for: (1) the Reactor Protection System, (2) the 
Isolation System, and (3) the Emergency Core Cooling System (ECCS). The 
deletion of instrumentation from the ECCS response time testing 
necessitates moving the remaining portion of the test to the ECCS 
system TS. In addition, the Note for Surveillance Requirement 
3.3.6.1.7, which reads: ``Radiation detectors may be excluded,'' is 
being removed since response time testing is not required for any 
radiation detector that provides a primary containment isolation signal 
as indicated in Table 3.3.6.1-1.
    Proposed TS Changes 1, 2, and 3 are supported by an analysis 
performed by the BWR Owners' Group (BWROG), with the licensee's 
participation. The analysis was submitted to the NRC for approval as 
Topical Report NEDO-32291, ``System Analyses for the Elimination of 
Selected Response Time Testing Requirements,'' Boiling Water Reactor 
Owners' Group, January 1994. The NRC approved the Topical Report by a 
Safety Evaluation Report (SER) issued on December 28, 1994, 
``Evaluation of Boiling Water Reactor Owners' Group Topical Report 
NEDO-32291, System Analyses for the Elimination of Selected Response 
Time Testing Requirements.'' The BWROG analysis demonstrates that other 
periodic tests required by TS, such as channel calibrations, channel 
checks, channel functional tests, and logic system functional tests, 
ensure that instrument response times are within acceptable limits. The 
applicability of the referenced analysis to Plant Hatch has been 
verified. Proposed Change 4 removes an unnecessary note, since no 
functions subject to this surveillance include radiation monitors.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

Basis for Proposed Changes 1, 2, and 3

    1. The changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated. The 
purpose of the proposed changes is to eliminate response time 
testing requirements for selected instrumentation in the RPS 
[Reactor Protection System], Isolation System], and ECCS. However, 
because of the continued application of other existing Technical 
Specifications requirements, such as channel calibrations, channel 
checks, channel functional tests, and logic system functional tests, 
the response time of these systems will be maintained within the 
acceptance limits assumed in plant safety analyses. This will assure 
successful mitigation of an initiating event. The proposed Technical 
Specifications changes do not affect the capability of the 
associated systems to perform their intended function within their 
required response time.
    The BWR Owners' Group (BWROG) has documented an evaluation in 
NEDO-32291, ``System Analyses for Elimination of Selected Response 
Time Testing Requirements,'' which was submitted to the NRC for 
review and approval as a Topical Report in January 1994 and 
subsequently approved by an NRC SER in December 1994. This 
evaluation demonstrates that response time testing is redundant to 
the other Technical Specifications requirements listed in the 
preceding paragraph. These other tests are sufficient to identify 
failure modes or degradation in instrument response time and ensure 
operation of the associated systems within acceptance limits. There 
are no known failure modes that can be detected by response time 
testing that cannot also be detected by the other Technical 
Specifications tests.
    2. The proposed changes will not create the possibility of a new 
or different kind of accident from any accident previously analyzed. 
As discussed above, the proposed Technical Specifications changes do 
not affect the capability of the associated systems to perform their 
intended function within the acceptance limits assumed in plant 
safety analyses.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety. The current Technical Specifications 
response times are based on the maximum allowable values assumed in 
the plant safety 

[[Page 35077]]
analyses, which conservatively establish the margin of safety. As 
described above, the proposed Technical Specifications changes do 
not affect the capability of the associated systems to perform their 
intended function within the allowed response time used as the basis 
for the plant safety analyses. Plant and system responses to an 
initiating event will remain in compliance with the assumptions of 
the safety analyses; therefore, the margin of safety is not 
affected.
    Although not explicitly evaluated, the proposed Technical 
Specifications changes enhance plant safety and operation by:
    a. Reducing the time safety systems are unavailable,
    b. Reducing safety system actuations,
    c. Reducing shutdown risk,
    d. Limiting radiation exposure to plant personnel, and
    e. Eliminating the diversion of key personnel to conduct 
unnecessary testing.
Basis for Proposed Change 4

    1. The change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated. The 
Note for SR 3.3.6.1.7 indicates that response time testing for 
radiation detectors that provide primary containment isolation 
signals as indicated in Table 3.3.6.1-1 is not required. However,
    Table 3.3.6.1-1 does not reference SR 3.3.5.1.7 for any 
radiation detector that provides primary containment isolation 
signals. The proposed change eliminates the potential for confusion 
during instrumentation surveillance testing. Deletion of the note 
will not prevent the radiation detectors from performing their 
intended function and will not affect the results of any accident 
analysis.
    2. The proposed changes will not create the possibility of a new 
or different kind of accident from any accident previously analyzed. 
As discussed above, the proposed Technical Specifications change 
eliminates the potential for confusion during instrumentation 
surveillance testing. This change does not modify any plant 
equipment or change any plant procedure that provides instructions 
for the operation of plant equipment. Therefore, the proposed change 
will not create the possibility of a new or different kind of 
accident from any previously analyzed.
    3. The proposed change does not involve a significant reduction 
in the margin of safety. The Note that is being deleted by the 
change states that testing is not required for instrument sensors 
which is not required by the SR. Therefore, the Note is superfluous 
and could cause confusion during instrumentation surveillance 
testing. The proposed change eliminates that potential. This change 
is conservative, since it deletes a statement that was intended to 
reduce the amount of surveillance testing performed on certain 
instrumentation. The proposed change does not affect plant 
equipment, procedures, or radiation release prevention and 
mitigating functions. Therefore, the proposed change does not 
involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513.

Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric 
Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-424 and 
50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, 
Georgia

    Date of amendment request: March 17, 1995.
    Description of amendment request: The amendments would revise 
Technical Specification (TS) 3.9.4, Containment Building Penetrations, 
to allow the personnel airlock to be open during core alterations or 
movement of irradiated fuel within the containment.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change to the Technical Specifications does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated. The proposed change to 
Specification 3.9.4 would allow the containment personnel airlock 
(PAL) to be open during fuel movement and core alterations. The PAL 
is currently closed during fuel movement and core alterations to 
prevent the escape of radioactive material in the event of a fuel 
handling accident. The PAL is not an initiator to any accident. 
Whether the PAL doors are opened or closed during fuel movement or 
core alterations has no effect on the probability of any accident 
previously evaluated.
    Allowing the PAL doors to be open during fuel movement and core 
alterations does increase the consequences of a fuel handling 
accident in the containment from essentially no offsite dose release 
to an estimated release of 65.6 rem to the thyroid and 0.28 rem to 
the whole body. However, the calculated offsite dose release is 
lower than the case analyzed in the FSAR [Final Safety Analysis 
Report] for an accident in the Spent Fuel Pool, with no filtration 
of the resulting release. In addition, the calculated doses are 
larger than the expected doses because the calculation does not 
incorporate the closing of the PAL door after the containment is 
evacuated. Closing the airlock door within 15 minutes results in a 
calculated offsite dose of 8.2 rem to the thyroid and 0.025 rem 
whole body. The projected dose to control room operators was 
reviewed and the projected dose remained below SRP acceptance limits 
as long as control room emergency ventilation was established within 
7 minutes. It was assumed the individual assigned to close the 
airlock doors remained stationed at the airlock for 15 minutes. A 
best estimate dose analysis indicated this individual could be 
expected to receive 5.6 rem to the thyroid and 0.15 rem whole body. 
The proposed change will significantly reduce the dose to other 
workers in the containment in the event of a fuel handling accident 
by speeding the containment evacuation process. The proposed change 
will also significantly decrease the wear on the PAL doors and, 
consequently, increase the availability of the PAL doors in the 
event of an accident.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change to the Technical Specifications does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated because the proposed change 
affects a previously evaluated accident, e.g., a fuel handling 
accident. It does not represent a significant change in the 
configuration or operation of the plant and, therefore, does not 
create the possibility of a new or different type of accident from 
any accident previously evaluated.
    3. The proposed change to the Technical Specifications does not 
involve a significant reduction in a margin of safety. The margin of 
safety as defined by 10 CFR Part 100 for a fission product release 
is 300 rem thyroid and 25 rem whole body for an individual exposed 
at the site boundary for two hours. The analysis shows values that 
are well below the acceptance limits. In fact, the margin remains 
essentially the same as previously evaluated by the NRC. There is no 
increase in calculated offsite dose resulting from a fuel handling 
accident. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.
    Based upon the preceding information, it has been determined 
that the proposed Technical Specifications addition does not involve 
a significant hazards consideration as defined by 10 CFR 50.92.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Burke County Public Library, 
412 Fourth Street, Waynesboro, Georgia 30830.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308.
    NRC Project Director: Herbert N. Berkow.

[[Page 35078]]


Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric 
Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-424 and 
50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, 
Georgia

    Date of amendment request: May 12, 1995.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) to support a one-time 
exemption from the requirement of Section III.D.1(a) of 10 CFR Part 50, 
Appendix J, and any other future Appendix J exemptions that may be 
approved by the NRC for Vogtle, Unit 1. Specifically, the TS change 
would insert the words ``Except as modified by NRC approved 
exemptions'' at the beginning of the first sentence of TS Surveillance 
Requirement 4.6.1.2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated. The 
proposed change does not involve a change to structures, systems, or 
components which would affect the probability of an accident 
previously evaluated in the Vogtle Electric Generating Plant (VEGP) 
Final Safety Analysis Report (FSAR). The change only provides a 
mechanism for implementing exemptions to 10 CFR 50, Appendix J 
containment leak rate testing criteria which have been approved by 
the NRC.
    2. The proposed change will not create the possibility of a new 
or different kind of accident from any accident previously analyzed. 
The amendment would not change the design, configuration, or method 
of plant operation. It only allows exemption to specific 10 CFR 50, 
Appendix J criteria as previously approved by the NRC.
    3. Operation of VEGP, Unit 1 in accordance with the proposed 
change will not involve a significant reduction in the margin of 
safety. The proposed change would not, in itself, change a safety 
limit, an LCO, or a surveillance requirement on equipment required 
for plant operation. Before the change could be used an exemption to 
10 CFR 50, Appendix J would have to be evaluated and approved by the 
NRC. The change only provides a way to implement NRC approved 
exemptions without violating the Technical Specifications.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Burke County Public Library, 
412 Fourth Street, Waynesboro, Georgia 30830.

GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania

    Date of amendment request: June 1, 1995.
    Description of amendment request: The proposed license amendment 
would revise the Technical Specifications (T.S.) for Three Mile Island 
Nuclear Station, Unit 1 (TMI-1) to delete the remaining portions of the 
TMI-1 Radiological Effluent Technical Specifications (RETS) and 
relocate them in accordance with the guidance contained in the Generic 
Letter 89-01 (GL 89-01) and NUREG-1430. The proposed change would also 
modify the Radiation Monitoring Systems surveillance requirements to 
specify only those radiation monitors that have Limiting Conditions for 
Operation (LCO), and revise some of the calibration frequencies.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence or the consequences of an accident 
previously evaluated. The proposed amendment allows relocation of 
the remaining RETS to the ODCM [Offsite Dose Calculation Manual] 
according to the guidance contained in GL 89-01 and NUREG-1430. This 
proposal simplifies the RETS, meets the regulatory requirements for 
radioactive effluent controls and radiological environmental 
monitoring, and is provided as a line-item improvement of the T.S.
    In addition, this proposed amendment specifies surveillance 
requirements only for those radiation monitors that have an LCO or 
specified operability requirements. The radiation monitors that are 
currently included in the T.S. surveillance program but have no 
associated LCO or specified operability requirement will be placed 
in the PM [preventive maintenance] program.
    Finally, the proposed amendment extends the interval between 
successive calibration surveillances for those radiation monitors 
evaluated herein. This change does not involve any change to the 
actual surveillance requirements, nor does it involve any change to 
the limits or restrictions on plant operations. The reliability of 
systems and components relied upon to prevent or mitigate the 
consequences of accidents previously evaluated is not degraded 
beyond that obtained from the currently defined quarterly interval. 
Assurance of system and equipment availability is maintained.
    This change does not involve any change to system or equipment 
configuration. Therefore, this change does not significantly 
increase the probability of occurrence or the consequences of an 
accident previously evaluated.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposal in part relocates procedural details, currently 
included in the T.S., on radioactive effluents to the ODCM. Future 
changes to these procedural details in the ODCM will be handled 
under the administrative controls for changes to the ODCM.
    In addition, this proposed amendment specifies surveillance 
requirements only for those radiation monitors that have an LCO or 
specified operability requirements. The radiation monitors that are 
currently included in the T.S. surveillance program but have no 
associated LCO or specified operability requirement will be placed 
in the PM program.
    Finally, the proposed amendment extends the interval between 
successive calibration surveillances for those radiation monitors 
evaluated herein. This change does not involve any change to the 
actual surveillance requirements, nor does it involve any change to 
the limits and restrictions on plant operations. This change does 
not involve any change to system or equipment configuration.
    Therefore, this change is unrelated to the possibility of 
creating a new or different kind of accident from any previously 
evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The procedural details being relocated to the ODCM are 
consistent with the guidance provided in GL 89-01 and NUREG-1430.
    In addition, this proposed amendment specifies surveillance 
requirements only for those radiation monitors that have an LCO or 
specified operability requirements. The radiation monitors that are 
currently included in the T.S. surveillance program but have no 
associated LCO or specified operability requirement will be placed 
in the PM program.
    Finally, the proposed amendment extends the interval between 
successive calibration surveillances for those radiation monitors 
evaluated herein. This change does not involve any change to the 
actual surveillance requirements, nor does it involve any change to 
the limits and restrictions on plant operations. The reliability of 
the radiation monitors is not significantly degraded beyond that 
obtained from the currently defined surveillance interval. Assurance 
of system availability is maintained.
    Therefore, it is concluded that operation of the facility in 
accordance with the proposed amendment does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this 

[[Page 35079]]
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Law/Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Phillip F. McKee.

Gulf States Utilities Company, Cajun Electric Power Cooperative, and 
Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit 
1, West Feliciana Parish, Louisiana

    Date of amendment request: May 30, 1995.
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TS) to increase the surveillance 
test period for the containment integrated leak rate test (ILRT) from 
40 plus or minus 10 months to every 10 years based on past performance. 
The change would also require testing on a more frequent basis if any 
test failures were to occur and to return to the 10 year period with 
subsequent performance improvements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) The proposed change does not involve a change to the plant 
design or operation. As a result, the proposed change does not 
affect any of the parameters or conditions that contribute to 
initiation of any accidents previously evaluated. Thus, the proposed 
change cannot increase the probability of any accident previously 
evaluated. The proposed change potentially affects the leak tight 
integrity of the containment structure designed to mitigate the 
consequences of a loss of coolant accident (LOCA). The function of 
the containment is to maintain functional integrity during and 
following the peak transient pressures and temperatures which result 
from any loss of coolant accident (LOCA). The containment is 
designed to limit fission product leakage following the design basis 
LOCA and analyses demonstrate that these offsite doses are less than 
those allowed under 10CFR100 design limits of 15 psig and 185 
deg.F. Because the proposed change does not alter the plant design, 
only the frequency of measuring containment leakage, the proposed 
change does not directly result in an increase in containment 
leakage. However, decreasing the test frequency can increase the 
probability that a large increase in containment leakage could go 
undetected for an extended period of time. These leakage paths 
include potential cracks in the containment structure and various 
penetrations through the containment structure. Based upon the 
results of the structural integrity test conducted as part of the 
preoperational or preservice test program and the periodic 
containment and drywell structural integrity surveillance tests, 
additional cracking of the containment is not expected during the 
remaining life to the plant. Ventilation and piping penetrations are 
designed with two isolation valves in series with one valve in the 
drywell and another either outside primary containment or in the 
wetwell. High energy lines that extend into the wetwell, such as the 
Main Steam and Feedwater lines, are encapsulated by guard pipes to 
direct energy to the drywell in case of a piping rupture.
    Electrical penetrations are sealed with a high strength/density 
material that will prevent leakage as well as provide radiation 
shielding. The TS ILRT acceptance criterion of 0.75 La [maximum 
allowable leakage rate at the calculated maximum accident pressure, 
Pa] provides margin for degradation. Containment performance 
data to date suggests that containment degradation, even during a 
ten (10) year interval between tests, will not exceed this margin.
    Based on the above, EOI [Entergy Operations, Inc.] has concluded 
that the proposed change will not result in a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    (2) The proposed change does not involve a change to the plant 
design or operation. As a result, the proposed change does not 
affect any of the parameters or conditions that could contribute to 
initiation of any accidents. This change involves the reduction in 
the Integrated Leak Rate Test frequency. The method of performing 
the test is not changed. No new accident modes are created by 
extending the testing intervals. No safety-related equipment or 
safety functions are altered as a result of this change. Extending 
the test frequency has no influence on, nor does it contribute to, 
the possibility of a new or different kind of accident or 
malfunction from those previously analyzed. Based upon the above, 
EOI has concluded that the proposed change will not create the 
possibility or a new or different kind of accident previously 
evaluated.
    (3) The proposed change only affects the frequency of measuring 
containment leakage and does not change the leakage rate limit. 
However, the proposed change can increase the probability that a 
large increase in containment leakage could go undetected for an 
extended period of time. Operational experience has shown that the 
leak tightness of the containment has been maintained significantly 
below the allowable leakage limit. In fact, an analysis was 
conducted to determine the potential risk to the public from the 
proposed change. Based on this analysis, under several different 
accident scenarios, the risk of radioactivity release from 
containment was found to be negligible.
    The margin of safety that has the potential of being impacted by 
the proposed change involves the offsite dose consequences of 
postulated accidents which are directly related to containment 
leakage rate. The containment isolation system is designed to limit 
leakage to La which is defined by the RBS Technical 
Specifications to be 0.26 percent by weight of the containment air 
per 24 hours at 7.6 psig (Pa). The limitation on containment 
leakage rate is designed to ensure that total leakage volume will 
not exceed the value assumed in the accident analyses at the peak 
accident pressure (Pa) or 7.6 psig.
    To provide additional conservatism, the measured overall 
integrated leakage rate is further limited to less than or equal to 
0.75 La during performance of the periodic Integrated Leak Rate 
Test and to less than or equal to 0.60 La (total combined 
leakage) for Type B and C leak rate tests. This is done to account 
for the possible degradation of the containment leakage barriers 
between tests. These acceptance criteria ensure that an acceptable 
margin of safety is being maintained and will not be altered by the 
proposed change. The preservation of this margin will continue to 
provide for potential degradation of the leakage barriers between 
tests. RBS [River Bend Station] presently has on docket with the 
staff a submittal (reference RBG-41133, Rev. 1 to LAR 93-14 dated 
January 18, 1995) that allows the acceptance criteria, between 
required leakage rate tests, to be less than or equal to 1.0 La 
since at less than or equal to 1.0 La, the offsite does 
consequences are bounded by the assumptions of safety analysis.
    No change in the method of testing is being proposed. The Type A 
test will continue to be done at full pressure (Pa) or greater. 
Primary containment penetrations which require Type B or C leak 
tests will be performed in the same manner as before. Other programs 
are in place to ensure that proper maintenance and repairs are 
performed during the service life of the primary containment and 
systems and components penetrating the primary containment.
    No change in the RBS allowable leakage rate is being proposed. 
These conservative leakage rates ensure that the containment leakage 
remains low. As a result, EOI has concluded that the proposed change 
will not result in a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, Louisiana 70803.
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 

[[Page 35080]]
    1400 L Street, N.W., Washington, DC 20005.
    NRC Project Director: William D. Beckner.

Gulf States Utilities Company, Cajun Electric Power Cooperative, and 
Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit 
1, West Feliciana Parish, Louisiana

    Date of amendment request: May 30, 1995.
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TS) to increase the time period 
for drywell leakage tests from eighteen months to five years based on 
performance. The new surveillance requirements would also reduce the 
time period if any failures occur and limit subsequent periods until 
drywell leakage test performance again improves.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) The proposed change does not involve a change to the plant 
design or operation. As a result, the proposed change does not 
affect any of the parameters or conditions that contribute to 
initiation of any accidents previously evaluated. Thus, the proposed 
change cannot increase the probability of any accident previously 
evaluated.
    The proposed change potentially affects the leak tight integrity 
of the drywell, a structure used to mitigate the consequences of a 
loss of coolant accident (LOCA). The function of the drywell is to 
channel the steam released from a LOCA through the suppression pool, 
limiting the amount of steam released to the primary containment 
atmosphere. This limits the containment pressurizations due to the 
LOCA. The leakage of the drywell is limited to ensure that the 
primary containment does not exceed its design limits of 185 deg.F 
and 15 psig. Because the proposed change does not alter the plant 
design, only the frequency of measuring the drywell leakage, the 
proposed change does not directly result in an increase in drywell 
leakage. However, decreasing the test frequency can increase the 
probability that a large increase in drywell bypass leakage could go 
undetected for an extended period of time. There are several 
potential sources of steam bypass leakage paths. These include 
potential cracks in the drywell concrete structure and various 
penetrations through the drywell structure. Based upon the results 
of the structural integrity test conducted as part of the 
preoperational or preservice test program, additional cracking of 
the drywell is not expected during the remaining life of the plant. 
Ventilation and piping penetrations are designed with two isolation 
valves in series with one valve in the drywell and another either 
outside primary containment or in the wetwell. High energy lines 
that extend into the wetwell, such as the Main Steam line and 
Feedwater lines, are encapsulated by guard pipe to direct energy to 
the drywell in case of a piping rupture. Electrical penetrations are 
sealed with a high strength/density material that will prevent 
leakage as well a provide radiation shielding. The TS DBLRT [Drywell 
Bypass Leakage Rate Tests] acceptance criterion of 10% of the design 
bypass leakage area parameter provides margin for degradation. 
Drywell performance data to date suggests that drywell degradation, 
even during a five year interval between tests, will not exceed this 
margin. RBS presently has on docket with the staff a submittal 
(reference EOI letter RBG-41133, Rev. 1 to LAR 93-14 dated January 
18, 1995) that allows the acceptance criteria, between required 
leakage rate tests, to be (bypass leakage area parameter) since at 
(bypass leakage area parameter) the containment temperature and 
pressurization response are bounded by the assumptions of the safety 
analysis.
    Based on the above, EOI has concluded that the proposed change 
will not result in a significant increase in the consequences of any 
accident previously evaluated.
    (2) The proposed change does not involve a change to the plant 
design or operation. As a result, the proposed change does not 
affect any of the parameters or conditions that could contribute to 
initiation of any accidents. Thus, the proposed change cannot create 
the possibility of an accident not previously evaluated.
    (3) The proposed change only affects the frequency of measuring 
the drywell bypass leakage rate and does not change the bypass 
leakage limit for the drywell. However, the proposed change can 
increase the probability that a large increase in drywell bypass 
leakage could go undetected for an extended period of time. 
Operational experience has shown that the leak tightness of the 
drywell has been maintained significantly below the allowable 
leakage limits. In fact, an analysis was conducted to determine the 
potential risk to the public from the proposed change. Based on this 
analysis, under several different accident scenarios, the risk of 
radioactivity release from containment was found to be negligible.
    As a result, EOI has concluded that the proposed change will not 
result in a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, Louisiana 70803.
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005.
    NRC Project Director: William D. Beckner.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, Michigan

    Date of amendment requests: May 25, 1995 (AEP:NRC:107IT).
    Description of amendment requests: The proposed amendments would 
implement a cycle- and burnup-dependent peaking factor penalty to the 
allowable power level. The Technical Specifications would be changed to 
refer to the Core Operating Limits Report for this burnup-dependent 
penalty.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Per 10 CFR 50.92, a proposed amendment will not involve a 
significant hazards consideration if the proposed amendment does 
not:
    (1) involve a significant increase in the probability or 
consequences of an accident previously evaluated,
    (2) create the possibility of a new or different kind of 
accident from any accident previously evaluated, or
    (3) involve a significant reduction in a margin of safety.

Criterion 1

    The proposed changes will not involve a significant increase in 
the probability of an accident previously evaluated because the 
changes will not result in a change to any of the process variables 
that might initiate an accident. There are no physical changes to 
the plant associated with this T/S change. The consequences of an 
accident previously evaluated will not be increased because the 
changes increase the penalty applied to FQ when it is measured 
to be increasing. FQ and allowable power level (APL) T/S 
surveillance requirements are not being changed. Furthermore, 
allowing a cycle and burnup dependent FQ penalty to be located 
in the COLR was accepted by the NRC in a [November 26, 1993] safety 
evaluation on WCAP-10216-P, Rev. 1 [``Relaxation of Constant Axial 
Offset Control- FQ Surveillance Technical Specification''].

Criterion 2

    The proposed changes will not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because the changes will involve no physical changes to the plant 
nor any changes in plant operations. Furthermore, the FQ and 
APL T/S surveillance requirements are not being changed, and the 
change to the FQ penalty is conservative.

Criterion 3

    The proposed amendment[s] will not involve a significant 
reduction in a margin of safety. When the increased FQ penalty 
is applied, it reduces the allowable power level, thus increasing 
the margin of safety.


[[Page 35081]]

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Project Director: Cynthia A. Carpenter, Acting.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, Michigan

    Date of amendment requests: May 25, 1995 (AEP:NRC:1124B).
    Description of amendment requests: The proposed amendments would 
modify the Technical Specifications (TS) to allow fuel reconstitution. 
The proposed change is a TS line item improvement per NRC Generic 
Letter 90-02, supplement 1, ``Alternative Requirements for Fuel 
Assemblies in the Design Features Section of Technical 
Specifications.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Per 10 CFR 50.92, a proposed change does not involve significant 
hazards consideration if the change does not:
    1. Involve a significant increase in the probability or 
consequence of an accident previously evaluated,
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated, or
    3. Involve a significant reduction in a margin of safety.

Criterion 1

    The proposed changes only modify the T/Ss such that 
reconstitution is recognized as acceptable under very limited 
circumstances. Reconstitution is limited to substitution of 
zirconium alloy or stainless steel filler rods, and must be in 
accordance with approved applications of fuel rod configurations. 
Although these changes permit reconstitution to occur without the 
need for a specific T/S change, an approved methodology is required 
prior to its application. Since the changes will allow substitution 
of filler rods for leaking or potentially leaking rods, the changes 
may actually reduce the radiological consequences of an accident. It 
is noted that the specific changes requested in this letter have 
previously been found acceptable by the NRC in GL 90-02 supplement 
1. For these reasons, we conclude that the changes will not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.

Criterion 2

    The proposed changes will not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because they will only affect the assembly configuration and can 
only be implemented in accordance with an NRC-approved methodology. 
The other aspects of plant design, operation limitations, and 
responses to events will remain unchanged. It is noted that the 
changes have previously been determined acceptable by the NRC in GL 
90-02 supplement 1.

Criterion 3

    The proposed amendment will not involve a significant reduction 
in a margin of safety because the changes can only be implemented in 
accordance with an NRC-approved methodology. It is noted that the 
changes have previously been determined acceptable by the NRC in GL 
90-02 supplement 1.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Project Director: Cynthia A. Carpenter, Acting.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, Michigan

    Date of amendment requests: May 25, 1995 (AEP:NRC:1200B).
    Description of amendment requests: The proposed amendments would 
modify the Technical Specifications to change the surveillance 
frequency of the manual actuation function for main steam line 
isolation. This change is consistent with the testing requirements for 
associated valves as specified in the American Society of Mechanical 
Engineers (ASME) Code Section XI inservice testing program at Cook.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Per 10 CFR 50.92, a proposed change does not involve significant 
hazards consideration if the change does not:
    1. Involve a significant increase in the probability or 
consequence of an accident previously evaluated,
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated, or
    3. Involve a significant reduction in a margin of safety.

Criterion 1

    This change will reduce the frequency of the surveillance 
testing on the MSIV [main steamline isolation valve] manual 
actuation circuitry from monthly to quarterly. Because of the risks 
involved in testing the dump valves, the reduction in test frequency 
may reduce the probability of an accidental unit trip and valve seat 
failure due to repeated cycling. Our review of the surveillance test 
history has shown that the system is highly reliable, and gives us 
confidence that the change in test frequency will not endanger 
public health and safety. Furthermore, the change to a quarterly 
surveillance interval is consistent with the testing performed for 
the dump valves per ASME Section XI. For these reasons, it is our 
belief that the proposed changes do not involve a significant 
increase in the probability or consequences of a previously 
evaluated accident.

Criterion 2

    The changes will not introduce any new modes of plant operation, 
nor will any physical changes to the plant be required. Thus, the 
changes should not create the possibility of a new or different kind 
of accident from any accident previously analyzed or evaluated.

Criterion 3

    This change will reduce the frequency of the surveillance 
testing on the MSIV manual actuation circuitry from monthly to 
quarterly. Our review of the surveillance test history has shown 
that the system is highly reliable, and gives us confidence that the 
change in test frequency will not endanger public health and safety. 
Furthermore, the change to quarterly surveillance is consistent with 
the testing performed for the dump valves per ASME Section XI. For 
these reasons, it is our belief that the proposed changes do not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Project Director: Cynthia A. Carpenter, Acting. 

[[Page 35082]]


Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, Michigan

    Date of amendment requests: May 26, 1995 (AEP:NRC:1210).
    Description of amendment requests: The proposed amendments would 
modify the Reactor Trip System Instrumentation and Engineered Safety 
Feature Actuation System Instrumentation sections of the Technical 
Specifications (TS) to relocate the tables of response time limits to 
the Updated Final Safety Analysis Report (UFSAR). These changes are a 
line item improvement of the TS in accordance with NRC Generic Letter 
93-08, ``Relocation of Technical Specification Tables of Instrument 
Response Time Limits.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Per 10 CFR 50.92, a proposed amendment will not involve a 
significant hazards consideration if the proposed amendment does 
not:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated,
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated, or
    (3) Involve a significant reduction in a margin of safety.

Criterion 1

    The proposed changes will not involve a significant increase in 
the probability of an accident previously evaluated because the 
changes will not result in a change to any of the process variables 
that might initiate an accident. There are no physical changes to 
the plant associated with the T/S change. The consequences of an 
accident previously evaluated will not be increased because the 
changes simply allow relocation of response time limits to the 
UFSAR. Time response testing will continue to be required by the T/
Ss. Any changes to the response time values will be made in 
accordance with the requirements of 10 CFR 50.59. It is noted that 
these T/S changes have previously been determined acceptable by the 
NRC in GL 93-08.

Criterion 2

    The proposed changes will not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because the changes will involve no physical changes to the plant 
nor any changes in plant operations. Time response testing will 
continue to be required by the T/Ss. Any changes to the time 
response values will be made in accordance with the requirements of 
10 CFR 50.59. It is noted that these changes have previously been 
determined acceptable by the NRC in GL 93-08.

Criterion 3

    The proposed amendment will not involve a significant reduction 
in a margin of safety because time response testing will continue to 
be required by the T/Ss. Any changes to the response time values 
will be made in accordance with the requirements of 10 CFR 50.59. It 
is noted that these changes have previously been determined 
acceptable by the NRC in GL 93-08.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Project Director: Cynthia A. Carpenter, Acting.

North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook 
Station, Unit No. 1, Rockingham County, New Hampshire

    Date of amendment request: May 30, 1995.
    Description of amendment request: The proposed amendment would 
change the upper limit for the moderator temperature coefficient (MTC) 
for certain operating conditions. Specifically, the upper limit 
specified in Technical Specification 3.1.1.3 for the MTC would be 
changed to +0.5 x 10-4 delta k/k/ deg.F for all rods out at the 
beginning of cycle for power levels up to 70% rated thermal power with 
a linear ramp to 0 delta k/k/ deg.F at 100% rated thermal power. The 
currently specified upper limit for all operating conditions is 0 delta 
k/k/ deg.F.
    A paragraph would be added to the Basis to Technical Specification 
3.1.1.3 providing a commitment to comply with the ATWS Rule and the 
basis for the Rule by assuring ATWS core damage frequency will remain 
below the Commission established target of 1.0 x 10-5 per reactor 
year. The commitment would be implemented by determining a more 
restrictive, cycle-specific upper MTC limit and placing it in the Core 
Operating Limits Report (COLR).
    Additionally, a reference for the analytical method used to 
determine the cycle-specific MTC upper limit would be added to TS 
6.8.1.6.b.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below.
    A. The changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated (10 CFR 
50.92(c)(1)). The proposed changes do not affect the manner by which 
the facility is operated and do not change any facility design feature 
or equipment which influences the initiation of an accident, therefore, 
there is no change in the probability of any accident previously 
analyzed. Each accident or transient, with the exception of the 
Anticipated Transient Without SCRAM (ATWS), has been analyzed for the 
proposed changes and has been approved previously by the Commission 
with the issuance of Amendment 33 (December 6, 1994) to the Facility 
Operating License. The proposed cycle-specific MTC to be included in 
the COLR will assure that the consequences of an ATWS will remain 
bounded by the analysis previously documented. Therefore, the 
consequences of previously evaluated accidents, including ATWS, will 
not be significantly increased by the proposed changes.
    B. The changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated (10 CFR 
50.92(c)(2)) because the changes proposed merely involve changes in the 
upper limits of MTC imposed by the Technical Specifications and COLR. 
No changes are made to the design or manner of operation of any 
structure, system or component and no new failure mechanisms are 
introduced.
    C. The changes do not involve a significant reduction in a margin 
of safety (10 CFR 50.92(c)(3)). The analyses of each accident or 
transient previously presented to support the issuance of Amendment 33 
were performed using the proposed upper MTC limit, and the results 
demonstrated that the acceptance criteria specified for each event are 
met. The cycle-specific MTC limit in the COLR will be adjusted to 
assure that the acceptance criteria for a postulated ATWS event are met 
thereby preserving the margin of safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration. 

[[Page 35083]]

    Local Public Document Room location: Exeter Public Library, 
Founders Park, Exeter, NH 03833.
    Attorney for licensee: Thomas Dignan, Esquire, Ropes & Gray, One 
International Place, Boston MA 02110-2624.
    NRC Project Director: Phillip F. McKee.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska

    Date of amendment request: May 31, 1995.
    Description of amendment request: The amendment would provide 
additional restrictions on the operation of the component cooling water 
(CCW) system heat exchangers to ensure that the CCW system temperature 
is maintained within its analyzed design basis.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    In preparation for, and in response to a service water system 
operational performance self assessment, the heat loads in the 
Component Cooling Water (CCW) system were reevaluated to determine 
the peak temperatures on the system and components cooled by the CCW 
system. It was determined that if all of the containment coolers 
were operating, the return temperature of the CCW system could 
exceed the 120 deg.F stated in the Updated Safety Analysis Report 
(USAR) as the maximum temperature of the system.
    During a Large Break Loss of Coolant Accident (LBLOCA) or a Main 
Steam Line Break Inside Containment (MSLB/IC), the containment air 
cooling units and containment air cooling and filtering units will 
automatically start to remove heat from the containment atmosphere. 
The heat sink for the containment air coolers is the CCW system. The 
heat removed from the containment atmosphere is transferred to the 
Raw Water (RW) system via the component cooling heat exchangers AC-
1A, B, C, and D. The heat is then ultimately rejected to the 
Missouri River by the RW system.
    Calculations indicate that the CCW return temperature (i.e., 
mixed exit temperature) from the component cooling heat exchangers 
could exceed 160 deg.F after a LBLOCA or MSLB/IC with the present TS 
minimum requirements for the heat exchangers. Further evaluation 
indicated that the CCW system (and components cooled by CCW) could 
withstand temperatures above the 120 deg.F temperature stated in the 
USAR, but a return temperature above 158 deg.F would require 
additional evaluation of thermal-induced stresses on the CCW return 
side pipe supports. In order to maintain the peak CCW return 
temperature to less than or equal to 158 deg.F, additional 
restrictions must be placed on the number of component cooling heat 
exchangers required to be operable.
    The current minimum requirements for component cooling heat 
exchangers are contained in Technical Specification (TS) 2.3, 
``Emergency Core Cooling System,'' and require that three of the 
four heat exchangers be operable when the plant is in operating 
Modes 1 and 2. Analyses show that three in service heat exchangers 
will maintain the CCW temperatures in an analyzed range following a 
DBA. In order to ensure that three heat exchangers are available, in 
conjunction with an assumed single failure, four are required to be 
operable. The proposed change would place additional restrictions on 
the operation of the CCW heat exchangers by requiring four heat 
exchangers to be operable in Modes 1 and 2, and if only three are 
operable then provide 14 days to restore the system to four operable 
heat exchangers.
    The proposed change does not involve a significant increase in 
the probability of an accident previously evaluated. The proposed 
change does not impact systems, structures, or components that are 
initiators of any analyzed accidents.
    The proposed change does not involve a significant increase in 
the consequences of an accident previously evaluated. The proposed 
change ensures that the CCW system and safety-related components 
cooled by the CCW will perform their safety functions in response to 
previously evaluated accidents. The proposed change was evaluated 
utilizing the probabilistic risk analysis model of the FCS 
Individual Plant Examination. The IPE concluded that the routine 
testing and maintenance activities, for the RW and CCW systems 
(e.g., inoperability of components for testing and maintenance) are 
not significant contributors to severe accident risk.
    Therefore, the proposed change would not increase the 
probability or consequences of an accident previously evaluated.
    (2) The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not create an initiator for a new or 
different kind of accident from those previously evaluated. The 
proposed change places additional restrictions on the operation of 
equipment to ensure that the CCW system and safety-related 
components cooled by the CCW will perform their safety functions. 
The additional restrictions were evaluated in combination with 
existing allowances on RW and CCW pump inoperability, to confirm 
that the peak CCW return temperature would be in an analyzed range, 
and will not adversely impact the operability of the CCW system or 
safety-related components cooled by CCW. These restrictions are 
valid up to and including a river temperature of 90 deg.F, which is 
the upper bound currently cited in the USAR.
    Various single active failures were postulated to determine the 
most limiting failure in conjunction with the maximum heat load from 
the containment air coolers. It was determined that with the river 
temperature less than 70  deg.F, a single failure of a RW valve to 
open on a component cooling heat exchanger would not raise the CCW 
return temperature to an unanalyzed level, but with the river 
temperature greater than or equal to 70  deg.F, the CCW return 
temperature could be at an unanalyzed level. Therefore, it is 
proposed that when the river temperature is greater than or equal to 
70  deg.F four heat exchangers have RW in service (i.e., RW valves 
open). Having RW in service eliminates the potential failure of a RW 
valve to auto-open as a credible single active failure.
    The proposed change ensures that the CCW system and safety-
related components cooled by the CCW will perform their safety 
functions. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    (3) The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change provides additional restrictions on the CCW 
system and ensures that the CCW system will perform its design 
safety function. These additional restrictions ensure that the CCW 
system will be capable of removing the maximum heat load from the 
containment cooling system following a DBA and thereby ensures that 
the containment pressure remains below its limit as assumed in the 
USAR. Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102.
    Attorney for licensee: James R. Curtiss, Winston & Strawn, 1400 L 
Street, N.W., Washington, DC 20005-3502.
    NRC Project Director: William H. Bateman.

Pennsylvania Power and Light Company, Docket No. 50-387, Susquehanna 
Steam Electric Station, Unit 1, Luzerne County, Pennsylvania

    Date of amendment request: May 5, 1995.
    Description of amendment request: This amendment would remove from 
the Susquehanna Steam Electric Station Unit 2 Technical Specifications, 
the listing of three residual heat removal (RHR) system valves in Table 
3.6.3-1, ``Primary Containment Isolation Valves'' These valves are no 
longer needed to support the steam condensing mode of the RHR system 
and are being removed from the plant during the Unit 2 seventh 

[[Page 35084]]
refueling and inspection outage in September of this year.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    With the prior deletion of the steam condensing mode of RHR and 
the isolation of the high and low pressure interfaces, the three 
pressure relief valves that are being removed from the plant have no 
active function. Their passive function of maintaining system or 
containment integrity will be fulfilled by blind flanges on 
equilvent. Also, the RHR and RCIC piping are provided with 
overpressure protection from other pressure relief valves. 
Therefore, the removal of these pressure relief valves does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The pressure relief valves that are being removed had two 
primary functions. First, they provided overpressure protection for 
the RHR and RCIC piping during the steam condensing mode of RHR. 
Since the steam condensing mode has been deleted from the plant, 
these valves no longer have that function. Also, overpressure 
protection of the RHR and RCIC piping is provided by other existing 
pressure relief valves. Second, these valves maintained system or 
containment integrity. When the pressure relief valves are removed 
from the plant, they will be replaced with blind flanges or 
equivalent that will maintain system or containment integrity. 
Therefore, the removal of the three pressure relief valves does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    Since the steam condensing mode of RHR has been eliminated, the 
three pressure relief valves have no active function. Their passive 
function of maintaining system or containment integrity will be 
fulfilled by blind flanges or equivalent. Also, overpressure 
protection of RHR and RCIC piping is provided by other existing 
pressure relief valves. Therefore, the removal of the three pressure 
relief valves does not involve a significant reduction in a margin 
of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701.
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz.

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of amendment request: May 19, 1995.
    Description of amendment request: The proposed Technical 
Specifications (TS) change would revise TS Table 3.3.3-3, ``Emergency 
Core Cooling System Response Times'' to reflect the value of 60 seconds 
for the High Pressure Coolant Injection system response time instead of 
30 seconds as currently specified.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed Technical Specifications (TS) change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The proposed TS change will increase the High Pressure Coolant 
Injection (HPCI) system response time from 30 seconds to 60 seconds. 
The proposed TS change does not involve any physical change in the 
plant configuration which may cause an accident, or affect safety-
related equipment performance or cause its failure. There is no 
increase in the consequences of an accident, because the HPCI 
response time increase does not affect the licensing basis Peak 
Cladding Temperature (PCT), which remains below the regulatory limit 
of 2200  deg.F.
    The Loss of Feedwater Flow (LOFW) event was evaluated for being 
potentially affected by the increased HPCI system response time. The 
HPCI system is one of the systems which provides reactor vessel 
water makeup inventory, and is initiated automatically on a low 
reactor water level (Level 2) signal. The LOFW analysis shows that 
Level 1 is not reached and that the top of the active fuel will 
remain covered throughout the event. Therefore, adequate core 
cooling will be maintained and no fuel damage will result. The 
probability of fuel failure will not be increased by this proposed 
TS change.
    Therefore, the proposed TS change does not involve an increase 
in the probability or consequences of an accident previously 
evaluated.
    2. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed TS change will increase the High Pressure Coolant 
Injection (HPCI) system response time from 30 seconds to 60 seconds. 
This proposed change is bounded by the current Emergency Core 
Cooling System (ECCS)--Loss-of-Coolant Accident (LOCA) analysis for 
Limerick Generating Station (LGS) Units 1 and 2. The change in HPCI 
system response time does not involve any physical modifications to 
the plant systems or equipment, nor does it introduce a new 
operational/failure mode, which might cause a different type of 
accident. In case of a Loss of Feedwater Flow (LOFW) event, the HPCI 
system will operate as designed, maintaining adequate core cooling.
    Therefore, the proposed TS change does not create the 
possibility of a new or different kind of accident, from any 
accident previously evaluated.
    3. The proposed TS change does not involve a significant 
reduction in a margin of safety.
    The following TS Bases were reviewed for potential reduction in 
the margin of safety:

3/4.5  Emergency Core Cooling System
2.1.4  Reactor Vessel Water Level

    The TS Bases do not discuss the High Pressure Coolant Injection 
(HPCI) system start time. The margin of safety, as defined in the TS 
Bases, will remain the same. The proposed TS change is in accordance 
with the current licensing basis Emergency Core Cooling System 
(ECCS)--Loss of Coolant Accident (LOCA) analysis for LGS Units 1 and 
2, and does not impact any safety limits of the plant. The HPCI 
system will operate as designed during the LOFW event, maintaining 
adequate core cooling.
    Therefore, the proposed TS change does not involve a reduction 
in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, Pennsylvania 19101.
    NRC Project Director: John F. Stolz.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 


[[Page 35085]]
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document rooms for 
the particular facilities involved.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2, 
and 3, Maricopa County, Arizona

    Date of application for amendments: December 7, 1994.
    Brief description of amendments: These amendments revise the Bases 
of TS 3/4.7.5, ``Ultimate Heat Sink'' (UHS), to describe the UHS as 
containing a 26-day supply of cooling water, instead of a 27-day 
supply. In addition, the reference to Regulatory Guide 1.27 in the 
bases of this TS would be revised to reference the January 1976 
revision rather than the March 1974 revision.
    Date of issuance: June 14, 1995.
    Effective date: June 14, 1995.
    Amendment Nos.: Unit 1--Amendment No. 93; Unit 2--Amendment No. 81; 
Unit 3--Amendment No. 64.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the associated Bases of the Technical 
Specifications.
    Date of initial notice in Federal Register: March 1, 1995 (60 FR 
11127) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 14, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power 
Station, Plymouth County, Massachusetts

    Date of application for amendment: February 9, 1995.
    Brief description of amendment: This amendment revises the reactor 
high water level trip level setting for the Group 1 isolation. The 
change will allow an increase to the main steam isolation valve high 
water level isolation setpoint.
    Date of issuance: June 15, 1995.
    Effective date: As of the date of issuance to be implemented within 
90 days.
    Amendment No.: 164.
    Facility Operating License No. DPR-35: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 15, 1995 (60 FR 
14017) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 15, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 
and 2, Will County, Illinois

    Date of application for amendments: May 20, 1994, as revised on 
February 2, 1995, and supplemented December 2, 1994, and March 14, 
1995.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TS) as they apply to Byron, Unit 1, and 
Braidwood, Unit 1, to incorporate an alternative repair criteria for 
defects found in the portion of the expanded steam generator tubes 
within the tubesheet.
    Date of issuance: June 22, 1995.
    Effective date: June 22, 1995.
    Amendment Nos.: 72, 72, 63, and 63.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: July 6, 1994 (59 FR 
34659) and March 29, 1995 (60 FR 16184). The Commission's related 
evaluation of the amendments is contained in a Safety Evaluation dated 
June 22, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481.

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 
1 and 2, Rock Island County, Illinois

    Date of application for amendments: September 15, 1992, as 
supplemented April 21, 1995.
    Brief description of amendments: This application upgrades the 
current custom Technical Specifications (TS) for Dresden and Quad 
Cities to the Standard Technical Specifications contained in NUREG-
0123, ``Standard Technical Specification General Electric Plants BWR/
4.'' This application upgrades only Sections 2.0 (Safety Limits and 
Limiting Safety System Settings), 3/4.11 (Power Distribution Limits), 
and 3/4.12 (Special Test Exceptions).
    Date of issuance: June 13, 1995.
    Effective date: Immediately, to be implemented no later than 
December 31, 1995, for Dresden Station and June 30, 1996, for Quad 
Cities Station.
    Amendment Nos.: 134, 128, 155, and 151.
    Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30. 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: May 10, 1995 (60 FR 
24906) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 13, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: for Dresden, Morris Area 
Public Library District, 604 Liberty Street, Morris, Illinois 60450; 
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon, 
Illinois 61021. 

[[Page 35086]]


Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 
1 and 2, Rock Island County, Illinois

    Date of application for amendments: December 15, 1993, as 
supplemented April 21, 1995.
    Brief description of amendments: These amendments upgrade the 
current custom Technical Specifications (TS) for Dresden and Quad 
Cities to the Standard Technical Specifications contained in NUREG-
0123, ``Standard Technical Specifications General Electric Plants BWR/
4.'' These amendments upgrade only Section 5.0 (Design Features). The 
amendments include the relocation of some requirements from the TS to 
licensee-controlled documents.
    Date of issuance: June 14, 1995.
    Effective date: Immediately, to be implemented no later than 
December 31, 1995, for Dresden Station and June 30, 1996, for Quad 
Cities Station.
    Amendment Nos.: 135, 129, 156, and 152
    Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30. 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: May 10, 1995 (60 FR 
24909) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 14, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: for Dresden, Morris Area 
Public Library District, 604 Liberty Street, Morris, Illinois 60450; 
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon, 
Illinois 61021.

Consumers Power Company, Docket No. 50-255, Palisades Plant, Van Buren 
County, Michigan

    Date of application for amendment: December 13, 1994, as 
supplemented May 3, 1995.
    Brief description of amendment: This amendment revises the 
Technical Specifications to add a high thermal performance (HTP) 
departure from nucleate boiling correlation to Safety Limit 2.1. The 
HTP correlation is used for HTP fuel loaded during recent fuel cycles.
    Date of issuance: June 13, 1995.
    Effective date: June 13, 1995.
    Amendment No.: 168.
    Facility Operating License No. DPR-20. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 10, 1995 (60 FR 
24910) The May 3, 1995, submittal provided clarifying information which 
was within the scope of the initial application and did not affect the 
staff's initial proposed no significant hazards considerations 
findings.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 13, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, Michigan 49423.

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver 
Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania

    Date of application for amendments: June 17, 1993, as supplemented 
October 20, 1993, and May 23, 1995.
    Brief description of amendments: These amendments revise the 
Appendix A technical specifications (TSs) for Unit 1 and Unit 2 by 
relocating the requirements of the radiological effluent technical 
specifications (RETS) and the solid radioactive wastes TSs from the 
Appendix A TSs to the offsite dose calculation manual (ODCM) or to the 
process control program (PCP) in accordance with the guidance provided 
in NRC Generic Letter 89-01 and NRC Report NUREG-1301. Programmatic 
controls are also being incorporated into the Administrative Controls 
section of the TSs. Additionally, editorial and definition changes are 
being made to facilitate the relocation of these requirements.
    Date of issuance: June 12, 1995.
    Effective date: June 12, 1995.
    Amendment Nos.: 188 and 70.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 4, 1993 (58 FR 
41504). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 12, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: May 15, 1995, as supplemented by letters 
dated May 19 and June 7, 1995.
    Brief description of amendment: The amendment was processed as an 
exigent amendment following issuance of a notice of enforcement 
discretion (NOED) by NRC letter dated May 17, 1995. The NOED and 
exigent technical specification (TS) amendment authorized the licensee 
to continue operating the reactor at power while the service water flow 
to the reactor building emergency coolers is less than the TS 
surveillance criteria.
    Date of issuance: June 9, 1995.
    Effective date: June 9, 1995.
    Amendment No.: 182.
    Facility Operating License No. DPR-51. Amendment revised the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: Yes (60 FR 27144, dated May 22, 1995). The notice 
provided an opportunity to submit comments on the Commission's proposed 
no significant hazards consideration determination. No comments have 
been received. The notice also provided for an opportunity to request a 
hearing by June 21, 1995, but stated that any such hearing would take 
place after issuance of the amendment. The Commission's related 
evaluation of the amendments, finding of exigent circumstances, and 
final determination of no significant hazards consideration is 
contained in a Safety Evaluation dated June 9, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: January 27, 1995.
    Brief description of amendment: The amendment changed the Appendix 
A Technical Specifications by increasing the allowable maximum 
enrichment for the spent fuel pool and containment temporary storage 
rack from 4.1 to 4.9 weight percent U-235 when fuel assemblies contain 
fixed poisons.
    Date of issuance: June 14, 1995.
    Effective date: June 14, 1995.
    Amendment No.: 108.
    Facility Operating License No. NPF-38. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 15, 1995 (60 FR 
14021)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 14, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.

[[Page 35087]]


Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie 
Plant, Unit No. 2, St. Lucie County, Florida

    Date of application for amendment: February 27, 1995.
    Brief description of amendment: This amendment will modify 
surveillance requirement (SR) 4.9.8.1 and 4.9.8.2 to allow a reduction 
in the required minimum shutdown cooling flow rate under certain 
conditions during operational MODE 6. In addition, the format of the SR 
will be changed to clarify the intent of the stated surveillances.
    Date of Issuance: June 14, 1995.
    Effective Date: June 14, 1995.
    Amendment No.: 76.
    Facility Operating License No. NPF-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 29, 1995 (60 FR 
16187) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 14, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of application for amendments: February 22, 1995.
    Brief description of amendments: The proposed changes eliminate 
reference to an automatic containment air lock tester from technical 
specification 4.6.1.3. The automatic air lock tester is no longer being 
used.
    Date of Issuance: June 22, 1995.
    Effective Date: June 22, 1995.
    Amendment Nos.: 137 and 77.
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 29, 1995 (60 FR 
16186) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 22, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Plant Units 3 and 4, Dade County, Florida

    Date of application for amendments: January 17, 1995.
    Brief description of amendments: These amendments concern 
implementation of Florida Power and Light nuclear physics methodology 
for calculations of the core operating limits report parameters.
    Date of issuance: June 9, 1995.
    Effective date: June 9, 1995.
    Amendment Nos. 174 and 168.
    Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 1, 1995 (60 FR 
11133) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 9, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199.

Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric 
Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-424 and 
50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, 
Georgia

    Date of application for amendments: October 3, 1994, as 
supplemented by letter dated March 1, 1995.
    Brief description of amendments: The amendments revise Technical 
Specification 3/4.4.9, Pressure/Temperature Limits, and its associated 
Bases, to provide new reactor coolant system heatup and cooldown 
limitations and new power-operated relief valve setpoints for the low 
temperature overpressure protection system.
    Date of issuance: June 8, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 87 and 65.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 21, 1994 (59 
FR 65814) The March 1, 1995, letter provided supporting technical data 
that did not change the scope of the October 1, 1994, application and 
initial proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated June 8, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Burke County Library, 412 
Fourth Street, Waynesboro, Georgia 30830.

GPU Nuclear Corporation, Docket No. 50-320, Three Mile Island Nuclear 
Station, Unit No. 2, (TMI-2), Dauphin County, Pennsylvania

    Date of application for amendment: October 9, 1991.
    Brief description of amendment: This amendment extends the 
expiration date of the license from November 9, 2009 to April 19, 2014.
    Date of issuance: June 21, 1995.
    Effective date: June 21, 1995.
    Amendment No.: 49.
    Possession-Only License No. DPR-73: The amendment extends the 
license expiration date.
    Date of initial notice in Federal Register: August 3, 1994 (59 FR 
39591). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 21, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, Walnut Street and Commonwealth 
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.

Gulf States Utilities Company, Cajun Electric Power Cooperative, and 
Energy Operations, Inc., Docket No. 50-458, River Bend Station, Unit 1, 
West Feliciana Parish, Louisiana

    Date of amendment request: February 22, 1994, as supplemented May 
19, 1995.
    Brief description of amendment: The amendment revised Technical 
Specifications 3.6.1.5, ``Main Steam--Positive Leakage Control 
System,'' and 3.6.1.10, ``Penetration Valve Leakage Control System,'' 
to add an allowed outage time of 7 days with both trains of each system 
inoperable. In addition, the allowed outage time for one train of the 
Penetration Valve Leakage Control System inoperable is increased from 7 
days to 10 days.
    Date of issuance: June 19, 1995.
    Effective date: June 19, 1995.
    Amendment No.: 80.
    Facility Operating License No. NPF-47. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 10, 1994 (59 FR 
11331) The additional information contained in the supplemental letter 
dated May 19, 1995, was clarifying in nature and thus, within the scope 
of the initial notice and did not affect the staff's proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated June 19, 1995. 

[[Page 35088]]

    No significant hazards consideration comments received. No.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, Louisiana 70803.

PECO Energy Company, Public Service Electric and Gas Company, Delmarva 
Power and Light Company, and Atlantic City Electric Company, Docket 
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Unit Nos. 2 
and 3, York County, Pennsylvania

    Date of application for amendments: November 17, 1994 as 
supplemented March 30, 1995.
    Brief description of amendments: The amendments change equipment 
designations, instrument range descriptions, instrument setpoints and 
surveillance requirements in the Peach Bottom Technical Specifications 
to reflect planned modifications to the main stack and vent stack 
radiation monitoring systems.
    Date of issuance: June 13, 1995.
    Effective date: June 13, 1995.
    Amendments Nos.: 204 and 207.
    Facility Operating License Nos. DPR-44 and DPR-56: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 15, 1995 (60 FR 
14027) The March 30, 1995, submittal provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 13, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.

PECO Energy Company, Public Service Electric and Gas Company, Delmarva 
Power and Light Company, and Atlantic City Electric Company, Docket 
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Unit Nos. 2 
and 3, York County, Pennsylvania

    Date of application for amendments: March 16, 1995.
    Brief description of amendments: These amendments change the 
existing Technical Specification requirements for source range neutron 
monitoring equipment while in the refueling mode to requirements based 
on NUREG-1433, ``Standard Technical Specifications General Electric 
Plants, BWR/4.''
    Date of issuance: June 13, 1995.
    Effective date: June 13, 1995.
    Amendments Nos.: 205 and 208.
    Facility Operating License Nos. DPR-44 and DPR-56: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 10, 1995 (60 FR 
24913) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 13, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.

PECO Energy Company, Public Service Electric and Gas Company, Delmarva 
Power and Light Company, and Atlantic City Electric Company, Docket No. 
50-277, Peach Bottom Atomic Power Station, Unit No. 2, York County, 
Pennsylvania

    Date of application for amendment: March 30, 1995, as supplemented 
by letter dated May 26, 1995.
    Brief description of amendment: The proposed amendment revises 
Technical Specification Section 4.7.D.1.b(1) by adding a footnote to 
exempt the High Pressure Coolant Injection motor-operated valve MO-2-
23-015 from quarterly stroke testing requirements until refueling 
outage 2RO11.
    Date of issuance: June 13, 1995.
    Effective date: June 13, 1995.
    Amendment No.: 206.
    Facility Operating License No. DPR-44: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 10, 1995 (60 FR 
24912) The May 26, 1995, submittal provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 13, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.

PECO Energy Company, Public Service Electric and Gas Company, Delmarva 
Power and Light Company, and Atlantic City Electric Company, Docket 
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Unit Nos. 2 
and 3, York County, Pennsylvania

    Date of application for amendments: March 22, 1995.
    Brief description of amendments: These amendments reduce the local 
leak rate test hold time specified in the Technical Specification 
Tables 3.7.2 through 3.7.4 from one hour to 20 minutes.
    Date of issuance: June 19, 1995.
    Effective date: June 19, 1995.
    Amendments Nos.: 207 and 209.
    Facility Operating License Nos. DPR-44 and DPR-56: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 10, 1995 (60 FR 
24913). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 19, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388 
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania

    Date of application for amendments: October 28, 1994, as 
supplemented by letter dated April 18, 1995.
    Brief description of amendments: These amendments delete, from the 
Technical Specifications, the surveillance and operability requirements 
for chlorine detection and the associated Bases as a result of the 
removal of bulk quantities of gaseous chlorine from the site.
    Date of issuance: June 19, 1995.
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment Nos.: 147 and 117.
    Facility Operating License Nos. NPF-14 and NPF-22. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 21, 1994 (59 
FR 65821). The April 18, 1995, letter provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination. 

[[Page 35089]]

    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 19, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701.

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of application for amendments: August 31, 1994.
    Brief description of amendments: This amendment revises the 
Technical Specifications to permit the operability requirement for the 
Feedwater/Main Turbine Trip System Actuation Instrumentation to be 
Operational Condition 1 greater than or equal to 25% Rated Thermal 
Power.
    Date of issuance: June 13, 1995.
    Effective date: June 13, 1995.
    Amendment Nos. 91 and 55.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 9, 1994 (59 FR 
55884) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 13, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of application for amendments: August 23, 1994.
    Brief description of amendments: Remove the 125/250 Vdc Class 1E 
Battery Load Cycle Table from the technical specifications (TS) and 
rephrase the surveillance requirements to be consistent with NUREG-
1433, ``Standard Technical Specifications'', and correct Amendments 71 
and 34, dated June 28, 1994, to change certain surveillance requirement 
intervals from 24 months to 18 months.
    Date of issuance: June 19, 1995.
    Effective date: June 19, 1995.
    Amendment Nos. 92 and 56.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 12, 1994 (59 FR 
51624) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 19, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of application for amendments: August 31, 1994.
    Brief description of amendments: These amendments relocate the 
requirements of TS 3/4.8.4.1, ``Primary Containment Penetration 
Conductor Overcurrent Protective Devices,'' to the Updated Final Safety 
Analysis Report and plant procedures.
    Date of issuance: June 22, 1995.
    Effective date: June 22, 1995.
    Amendment Nos. 93 and 57.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 9, 1994 (59 FR 
55884) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 22, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of application for amendments: August 12, 1994, as 
supplemented by letter dated March 29, 1995.
    Brief description of amendments: These amendments revise the action 
statements regarding emergency core cooling systems to allow continued 
operation in the event that the high pressure coolant injection system, 
one core spray subsystem and/or one low pressure coolant injection 
subsystem are inoperable.
    Date of issuance: June 22, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos. 94 and 58.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 12, 1994 (59 FR 
51623). The March 29, 1995, letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 22, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of application for amendments: August 31, 1994.
    Brief description of amendments: The amendments permit the 
operability of one Low Pressure Coolant Injection subsystem of Residual 
Heat Removal while the subsystem is aligned and operating in the 
Shutdown Cooling Mode during Operational Conditions (OPCONs) 4 and 5.
    Date of issuance: June 22, 1995.
    Effective date: June 22, 1995.
    Amendment Nos. 95 and 59.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 9, 1994 (59 FR 
55884). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 22, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311, 
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New 
Jersey

    Date of application for amendments: November 18, 1994.
    Brief description of amendments: The amendments revise the 
Reactivity Control System Technical Specification Limiting Conditions 
for Operation for boration flow paths and charging pumps by reducing 
the number of operable charging pumps required for boron addition in 
Mode 4 from two to one.
    Date of issuance: June 12, 1995.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment Nos. 169 and 151.
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications. 

[[Page 35090]]

    Date of initial notice in Federal Register: January 4, 1995 (60 FR 
505). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 12, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079.

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311, 
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New 
Jersey

    Date of application for amendments: June 29, 1994, as supplemented 
August 8, 1994, and May 2, 1995.
    Brief description of amendments: The amendments increase the 
Technical Specification minimum volume of emergency diesel generator 
fuel oil contained in the Diesel Fuel Oil Storage Tanks at both units 
of the Salem station.
    Date of issuance: June 20, 1995.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment Nos. 170 and 152.
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 17, 1994 (59 FR 
42346). The August 8, 1994, and May 2, 1995, letters provided 
clarifying information that did not change the initial proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 20, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079.

South Carolina Electric & Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 
No. 1, Fairfield County, South Carolina

    Date of application for amendment: March 6, 1995, as supplemented 
on May 5, 1995 and June 6, 1995.
    Brief description of amendment: The amendment deletes a license 
condition that required the licensee to maintain a seismic monitoring 
network around the Monticello Reservoir.
    Date of issuance: June 13, 1995.
    Effective date: June 13, 1995.
    Amendment No.: 124.
    Facility Operating License No. NPF-12. Amendment revises the 
operating license.
    Date of initial notice in Federal Register: March 29, 1995 (60 FR 
16201). The May 5, 1995 and June 6, 1995 submittals provided 
supplemental information that did not change the initial proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated June 13, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Fairfield County Library, 300 
Washington Street, Winnsboro, SC 29180.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama

    Date of application for amendments: November 15, 1994; superseded 
March 7, 1995 (TS 350).
    Brief Description of amendment: The amendments remove the 
frequencies specified in the Technical Specifications for performing 
audits and delete the requirement to perform the Radiological Emergency 
Plan, Physical Security Plan, and Safeguard Contingency Plan reviews.
    Date of issuance: June 19, 1995.
    Effective Date: June 19, 1995.
    Amendment Nos.: 221, 236 and 195.
    Facility Operating License Nos. DPR-33, DRP-52 and DPR-68: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: December 21, 1994 (59 
FR 65823); superseded March 29, 1995 (60 FR 16202). The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated June 19, 1995.
    No significant hazards consideration comments received: None.
    Local Public Document Room Location: Athens Public Library, South 
Street, Athens, Alabama 35611.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendments: April 6, 1995 (TS 95-02).
    Brief description of amendments: The amendments add a limiting 
condition for operation that allows equipment to be returned to service 
under administrative control to perform operability testing and 
establishes the time interval to place an inoperable channel in the 
bypass condition.
    Date of issuance: June 13, 1995.
    Effective date: June 13, 1995.
    Amendment Nos.: 202 and 192.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: April 26, 1995 (60 FR 
20530). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 13, 1995.
    No significant hazards consideration comments received: None.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendments: April 6, 1995 (TS 95-05).
    Brief description of amendments: The amendments revise the 
technical specifications by deleting Tables 3.6-1, 3.6-2, and 3.8-2 and 
referenced to them, incorporating related guidance and justification, 
and modifying the specification related to electrical equipment 
protective devices in accordance with Generic Letter 91-08.
    Date of issuance: June 13, 1995.
    Effective date: June 13, 1995.
    Amendment Nos.: 203 and 193.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: May 10, 1995 (60 FR 
24919). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 13, 1995.
    No significant hazards consideration comments received: None.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendments: April 6, 1995 (TS 95-06).
    Brief description of amendments: The amendments remove the 
technical specification requirements related to crane travel over the 
spent fuel pool.
    Date of issuance: June 14, 1995.
    Effective date: June 14, 1995.
    Amendment Nos.: 204 and 194.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: April 26, 1995 (60 FR 
20529). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 14, 1995. 

[[Page 35091]]

    No significant hazards consideration comments received: None.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont 
Yankee Nuclear Power Station, Vernon, Vermont

    Date of application for amendment: October 28, 1994.
    Brief description of amendment: The amendment removes the Neutron 
Monitoring System and Control Rod Position instrumentation from the 
Vermont Yankee Technical Specifications for post-accident monitoring 
and incorporates administrative changes.
    Date of issuance: June 20, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 145.
    Facility Operating License No. DPR-28. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 10, 1995 (60 FR 
24922). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 20, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia.

    Date of application for amendments: June 9, 1994.
    Brief description of amendments: These amendments modify the 
Chemical and Volume Control System and Safety Injection System 
Technical Specifications.
    Date of issuance: May 31, 1995.
    Effective date: May 31, 1995.
    Amendment Nos. 199 and 199.
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 20, 1994 (59 FR 
37089). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 31. 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185.

    Dated at Rockville, Maryland, this 27th day of June 1995.

    For the Nuclear Regulatory Commission.
Jack W. Roe,
Director, Division of Reactor Projects--III/IV, Office of Nuclear 
Reactor Regulation.
[FR Doc. 95-16249 Filed 7-3-95; 8:45 am]
BILLING CODE 7590-01-P